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

NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility

Operating Licenses Involving no Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from March 17, 1997 through March 28, 1997. The

last biweekly notice was published on March 26, 1997 (62 FR 14457).

Notice of Consideration of Issuance of Amendments to Facility Operating

Licenses, Proposed no Significant Hazards Consideration Determination,

and Opportunity for a Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

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

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

Involve a significant increase in the probability or consequences of an

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

or different kind of accident from any accident previously evaluated;

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

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

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

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

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

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

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

the Commission may issue the license amendment before the expiration of

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

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

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

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

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

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

Review and Directives Branch, Division of Freedom of Information and

Publications Services, Office of Administration, U.S. Nuclear

Regulatory Commission, Washington, DC 20555-0001, and should cite the

publication date and page number of this Federal Register notice.

Written comments may also be delivered to Room 6D22, Two White Flint

North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15

p.m. Federal workdays. Copies of written comments received may be

examined at the NRC Public Document Room, the Gelman Building, 2120 L

Street, NW., Washington, DC. The filing of requests for a hearing and

petitions for leave to intervene is discussed below.

By May 9, 1997, the licensee may file a request for a hearing with

respect to issuance of the amendment to the subject facility operating

license and any person whose interest may be affected by this

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

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

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

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

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

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

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

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

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

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

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

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

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

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert

[[Page 17224]]

opinion which support the contention and on which the petitioner

intends to rely in proving the contention at the hearing. The

petitioner must also provide references to those specific sources and

documents of which the petitioner is aware and on which the petitioner

intends to rely to establish those facts or expert opinion. Petitioner

must provide sufficient information to show that a genuine dispute

exists with the applicant on a material issue of law or fact.

Contentions shall be limited to matters within the scope of the

amendment under consideration. The contention must be one which, if

proven, would entitle the petitioner to relief. A petitioner who fails

to file such a supplement which satisfies these requirements with

respect to at least one contention will not be permitted to participate

as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

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

and to the attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

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

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

Carolina

Date of amendments request: March 5, 1997.

Description of amendments request: The proposed amendments would

incorporate a new Technical Specification (TS) for instrumentation

associated with automatic isolation of a pathway for release of non-

condensible gases from the main condenser. At power levels of 5 percent

or less, mechanical vacuum pumps are used to remove non-condensible

gases from the condenser using a pathway to the release stack that

bypasses the normal holdup and filter train. The proposed TS will

require that four channels of the main steam line radiation--high

isolation function be capable of tripping the mechanical vacuum pumps

and closing an isolation valve in the release pathway. Surveillance

requirements are included in the TS to ensure the isolation

instrumentation will perform its intended function.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendments do not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed change incorporates a new Technical Specification

3/4.3.8, ``Condenser Vacuum Pump Isolation Instrumentation.'' This

specification will require that the main steam line radiation--high

isolation function be capable of tripping the condenser vacuum

pump(s) and isolate the associated common isolation valve. Four

instrumentation channels of this function are required to be

operable when the unit is in OPERATIONAL CONDITION 1 or 2 with a

condenser vacuum pump in operation. Adding the requirement to trip

the condenser vacuum pumps does not affect the probability of an

accident previously evaluated. The probability of component failure

of the proposed design for condenser vacuum pump isolation devices

is the same as that of the original licensing basis. As a result,

the capability to isolate the condenser vacuum pump will not be

significantly impacted.

CP&L contracted Scientech-NUS to recalculate the main control

room doses resulting from a control rod drop accident assuming main

steam line radiation monitors isolate the condenser vacuum pump(s)

and determined the dose to be 23.2 rem thyroid and 0.05 rem whole

body, which is less than the General Design Criterion (GDC) 19/

Standard Review Plan (SRP) Section 6.4 limits of 30 rem thyroid and

5 rem whole body. The offsite doses at the exclusion area boundary

after 2 hours are 0.16 rem thyroid and 0.015 rem whole body, which

is less than the SRP Section 15.4.9 limits. The low population zone

(LPZ) dose is estimated to be about 1 rem thyroid, which is also

well below regulatory limits. Therefore, the proposed [amendments

do] not increase the consequences of an accident previously

evaluated.

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

a new or different kind of accident from any accident previously

evaluated.

The proposed [amendments add] new requirements to ensure the

capability to trip the condenser vacuum pump(s). The proposed

[changes do] not affect the operability of equipment designed to

mitigate the consequences of an accident nor [do they] create a

potential to initiate a new type 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 proposed license [amendments do] not involve a

significant reduction in a margin of safety.

The safety-related main steam line radiation monitors provide a

highly reliable means to detect radioactivity resulting from a

control rod drop accident and will provide automatic trip of the

condenser vacuum pumps and isolation of the associated isolation

valve. Use of the main steam line radiation monitors for this

application is consistent with the original Brunswick Steam Electric

Plant design for condenser pump and associated valve isolation. CP&L

contracted Scientech-NUS to recalculate the main control room doses

resulting from a control rod drop accident assuming main steam line

radiation monitors isolate the condenser

[[Page 17225]]

vacuum pump(s) and determined it to be 23.2 rem thyroid and 0.05 rem

whole body, which is less than the GDC 19/SRP Section 6.4 limits of

30 rem thyroid and 5 rem whole body. The offsite doses at the

exclusion area boundary after 2 hours are 0.16 rem thyroid and 0.015

rem whole body, which is less than the SRP Section 15.4.9 limits.

LPZ dose is estimated to be about 1 rem thyroid, which is also well

below regulatory limits. 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: University of North Carolina

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

Wilmington, North Carolina 28403-3297.

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

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

Raleigh, North Carolina 27602.

NRC Project Director: Mark Reinhart, Acting.

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: February 18, 1997.

Description of amendment request: The proposed change revises the

Plant System Turbine Cycle Technical Specification (TS) 3/4.7.1 by

revising the power range high neutron flux setpoint values in TS Table

3.7-1.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The high flux setpoints are being revised to provide additional

margin against secondary side overpressurization for LOL/TT [loss-

of-load/turbine trip] events. The proposed revision will not create

any loss or reduction in redundancy or diversity in the reactor

protection systems that would increase the probability of a

previously evaluated accident. The high flux setpoints are being

revised to ensure that the consequences of a previously evaluated

accident do not increase.

Therefore, there would be no increase in the probability or

consequences of an accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

No new or previously unanticipated failure mechanisms are

introduced by the proposed change. No new failure modes have been

created by the proposed change. No new credible event or initiating

factor is introduced. Reactor power is limited to ensure that the

secondary system is not overpressurized.

Therefore, the proposed change does not create the possibility

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

evaluated.

3. The proposed amendment does not involve a significant

reduction in the margin of safety.

The margin of safety as defined in the basis of the Technical

Specification does not decrease. This change is proposed to ensure

that the secondary system pressure will be limited to within 110% of

its design pressure during the most severe anticipated operational

transient. The revised high flux setpoints are intended to bound the

allowable operating configurations of TS Table 3.7-1.

Therefore, the proposed change does not involve a significant

reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Cameron Village Regional

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

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

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

Raleigh, North Carolina 27602.

NRC Project Director: Mark Reinhart, Acting.

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: February 21, 1997.

Description of amendment request: The proposed change adds a

definitive time limit to Technical Specification 3.3.2 in Action 16 of

Table 3.3-3 to place an inoperable channel into 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. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed change does not affect the operation or design of

the plant in any way. The requirement to place the channel into

bypass already exists and this change simply provides a specific

time limit. This logic circuit is not an initiator of any event and

with no change in logic or operation there is no change in

consequences.

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

new or different kind of accident from any accident previously

evaluated.

The proposed specific time limit does not involve any physical

alterations or additions to plant equipment or alter the manner in

which any safety-related system performs its function. Therefore,

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

different kind of accident from any accident previously evaluated.

3. The proposed amendment does not involve a significant

reduction in the margin of safety.

The proposed change replaces an indeterminate time period with a

specific limit of six hours. Six hours is a reasonable period in

which to complete this requirement and is identical to the time

allowed for these functions in NUREG-1431 [Standard Specifications

Westinghouse Plants]. Therefore, the proposed change does not

involve a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Cameron Village Regional

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

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

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

Raleigh, North Carolina 27602.

NRC Project Director: Mark Reinhart, Acting.

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: February 27, 1997.

Description of amendment request: The proposed change adds sleeve

installation as an alternative to tube plugging for repairing degraded

steam generator tubes to Technical Specification 3/4.4.5, Steam

Generators.

[[Page 17226]]

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The only equipment affected by sleeving is the steam generator

tubes. The most severe malfunction of a steam generator tube is a

tube rupture. The consequences of a ruptured sleeve are no greater

than the consequences of a ruptured tube. Sleeving does not increase

the probability of a steam generator tube failure because the

sleeved tube has been shown to have a significant safety factor for

burst and collapse pressures as well as demonstrated acceptable

resistance to corrosion and fatigue loading. Thus, a steam generator

with sleeved tubes would perform in the same manner as one without

sleeved tubes.

A sleeved tube is functionally equivalent to an unsleeved tube

except for less effective heat transfer due to the air gap and a

slightly higher pressure drop due to the primary flow restriction.

These differences are bounded by the current tube plugging limits.

Analysis and testing have demonstrated that the sleeves are

structurally adequate to withstand the load existing within the

steam generator tubes whether the original tube is still intact or

is breeched.

There is no increase in the possibility for increased fatigue

loadings. There is no possibility for the sleeve to become dislodged

from its plugging location and enter the RCS [Reactor Coolant

System] flow path.

The plant safety analysis for tube plugging bounds tube

sleeving.

The proposed change has no significant effect on the

configuration of the plant. The proposed change does not affect the

way in which the plant is operated. Therefore, there would be no

increase in the probability or consequences of an accident

previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

A sleeved tube is functionally equivalent to an unsleeved tube

except for less effective heat transfer due to the air gap and a

slightly higher pressure drop due to the primary flow restriction.

These differences are bounded by the current tube plugging limits.

The sleeved tube has been shown to have a significant safety

factor for burst and collapse pressures as well as demonstrated

acceptable resistance to corrosion and fatigue loading. Thus, a

steam generator with sleeved tubes would perform in the same manner

as one without sleeved tubes.

The proposed change has no significant effect on the

configuration of the plant. The proposed change does not affect the

way in which the plant is operated. Therefore, the proposed change

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

accident from any accident previously evaluated.

3. The proposed amendment does not involve a significant

reduction in the margin of safety.

The proposed revision to permit the installation of tube sleeves

does not reduce the margin of safety as presently defined in

Technical Specification BASES section 3/4.4.5. This margin of safety

includes primary to secondary leakage limits and tube plugging

limits which are not changed by the proposed amendment. The analyses

and testing of the proposed sleeve design demonstrates that the

structural integrity of the RCS is maintained. Design of the tube

sleeve considers mechanical/structural aspects, water chemistry and

metallurgical aspects as well as thermal/hydraulic considerations.

Therefore, the proposed change does not involve a significant

reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Cameron Village Regional

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

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

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

Raleigh, North Carolina 27602.

NRC Project Director: Mark Reinhart, Acting.

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: March 10, 1997.

Description of amendment request: The proposed changes to Technical

Specification 3.5.1 provide an optional method of meeting surveillance

requirements by allowing the use of instrument readings to meet

surveillance 4.5.1.1.a.1, and adds a new Action c to cover a condition

in which one accumulator has a boron concentration not within limits.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The accumulators are not initiators of any event and so the

probability of occurrence of an event is unaffected by either of the

proposed changes. The use of actual instrumentation readings to

comply with the surveillance does not change the function or

performance of the accumulators and thus does not affect any

accident consequences. The increase in the allowed time to restore

the boron concentration to within limits is consistent with allowed

out of service times for other Emergency Safeguards equipment.

It will not have a significant impact on subcriticality during

reflood. Therefore, there will be no increase in the consequences of

an accident.

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

new or different kind of accident from any accident previously

evaluated.

The proposed changes to the accumulator specification do not

involve any physical alterations or additions to plant equipment or

alter the manner in which any safety-related system performs its

function. Therefore, the proposed change does not create the

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

previously evaluated.

3. The proposed amendment does not involve a significant

reduction in the margin of safety.

The proposed change to the surveillance requirement provides an

equivalent means of meeting the requirement. Since there is no

change in either the accumulator limits or the surveillance

frequency, there is no reduction in safety margin. The new Action c

to address returning the boron concentration of a single accumulator

to within limits allows an out of service time commensurate with the

times allowed for other Engineered Safeguards Features. The boron

concentration of one accumulator does not have a significant impact

on subcriticality during reflood and thus does not involve a

reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Cameron Village Regional

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

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

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

Raleigh, North Carolina 27602.

NRC Project Director: Mark Reinhart, Acting.

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: March 14, 1997.

Description of amendment request: The amendment will revise the

Final Safety Analysis Report to include the

[[Page 17227]]

evaluation of a spent fuel cask drop analysis.

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 changes described do not impact the probability of

occurrence of accidents previously analyzed. Removal of the valve

box covers and all but four of the cask closure head sleeve nuts has

no impact on accident initiators. Dose assessments using maximum

potential releases assuming failure of the spent fuel and

radionuclide release through the gap between the cask closure head

and the cask or damage to the valves show that no significant

increase in consequences of an accident previously evaluated would

occur. [Therefore, the proposed change 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.

Compromising the integrity of the cask by removing the valve box

covers and closure head sleeve nuts in preparation for unloading the

spent fuel from the cask does not create the possibility of a new

type of accident or equipment malfunction. No safety-related

equipment, safety function, or operations of plant equipment will be

altered as a result of this change. Therefore, the proposed changes

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

from any accident previously evaluated.

3. The proposed change does not involve a significant reduction

in the margin of safety.

The NRC basis for acceptance of a spent fuel cask drop is

documented in Section 15.7.5 of the Safety Evaluation Report, NUREG-

1038, dated November 1983. It states, ``* * * no loss of cask

integrity is postulated to occur in the event of a drop, and the

staff concludes there will be no significant radiation released to

the environment. The radiological consequences will be less than a

small fraction of the 10 CFR 100 exposure guideline values.''

As described in the proposed change, even though complete cask

integrity may not be preserved in the event of a loaded cask drop

with the valve box covers removed or with only four, rather than 32,

closure head sleeve nuts installed, the radiological consequences

calculated using conservative assumptions were determined to be a

small fraction of the 10 CFR 100 values. Therefore, the proposed

change does not involve a significant reduction in the margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Cameron Village Regional

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

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

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

Raleigh, North Carolina 27602.

NRC Project Director: Mark Reinhart, Acting.

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

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

Date of amendment request: June 20, 1996, as supplemented by

letters dated December 30, 1996, and March 5, 1997.

Description of amendment request: The proposed amendments would

change the Technical Specifications (TS) by incorporating NRC approved

thermal limit licensing methodology in the list of approved

methodologies used in establishing the fuel cycle specific thermal

limits. In addition, the proposed amendment would correct mirror

editorial items in the TS.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The probability of an evaluated accident is derived from the

probabilities of the individual precursors to that accident. The

consequences of an evaluated accident are determined by the

operability of plant systems designed to mitigate those

consequences. Limits will be established consistent with NRC

approved methods to ensure that fuel performance during normal,

transient and accident conditions is acceptable. The proposed

Technical Specifications amendment reflects NRC approved SPC

methodology used to analyze normal operations, including anticipated

operational occurrences (AOOs), and to determine the potential

consequences of accidents.

Licensing Methods and Models

The proposed amendment is to support operation with NRC approved

fuel and licensing methods supplied from Siemens Power Corporation

[SPC]. In accordance with [Updated Final Safety Analysis Report]

UFSAR Chapter 15, the same accidents and transients will be analyzed

with the new fuel and methods. The latest NRC approved revision to

the Siemens [loss-of-coolant accident] LOCA analysis methodology

(Reference: ANF-91-048(P)(A), Advanced Nuclear Fuels Corporation

Methodology for Boiling Water Reactors EXEM BWR Evaluation Model)

will be used to evaluate the ATRIUM-9B and other co-resident fuel

types. The other licensing analysis methods and models are also NRC

approved. The approved methods and models are used to determine the

fuel thermal limits (e.g., average planar linear heat generation

rate, transient linear heat generation rate, minimum critical power

ratio and linear heat generation rate). The SPC core monitoring code

enables the site to monitor keff as well as control rod density

to perform the reactivity anomaly surveillance. Therefore, the

change in licensing analysis methods and models does not

significantly increase the probability of an accident or the

consequences of an accident previously identified. The support

systems for minimizing the consequences of transients and accidents

are not affected by the proposed amendment.

New Fuel Design

The use of reload quantities of ATRIUM-9B fuel at Dresden does

not involve a significant increase in the probability or

consequences of any accident previously evaluated in the [Final

Safety Analysis Report] FSAR. The ATRIUM-9B fuel is generically

approved for use as a reload BWR fuel type (Reference: ANF-89-

014(P)(A) Revision 1 Supplement 1, Generic Mechanical Design for

Advanced Nuclear Fuels 9X9-IX and 9X9-9X BWR Reload Fuel). Limiting

postulated occurrences and normal operation have been analyzed using

NRC-approved methods for the ATRIUM-9B fuel design to ensure that

safety limits are protected and that acceptable transient and

accident performance is maintained.

The reload fuel has no adverse impact on the performance of in-

core neutron flux instrumentation or CRD response. The ATRIUM-9B

fuel design will not adversely affect performance of neutron

instrumentation nor will it adversely affect the movement of control

blades relative to the current Dresden fuel type, the Siemens

manufactured 9x9-2. The exterior dimensions of the ATRIUM-9B fuel

have been evaluated by ComEd; the ATRIUM-9B fuel design provides

adequate clearances relative to the co-resident 9x9-2 fuel. Thus, no

increased interactions with the adjacent control blade or nuclear

instrumentation are created. Additionally, given the above mentioned

overall envelope similarities, no problems are anticipated with

other station equipment such as the fuel storage racks, the new fuel

inspection stand and the spent fuel storage pool fuel preparation

machine. Therefore, the probability of adverse interactions between

the ATRIUM-9B fuel and components in the core and fuel handling

equipment is not significantly increased.

The ATRIUM-9B design is neutronically compatible with the

existing fuel types and core components in the Dresden core. SPC

tests have demonstrated that the ATRIUM-9B fuel design is

hydraulically compatible with the co-resident 9x9-2 fuel. The bundle

pressure drop characteristics of the ATRIUM

[[Page 17228]]

9B bundle are similar to those of the 9x9-2 fuel design, hence core

thermal-hydraulic stability characteristics are not adversely

affected by the ATRIUM-9B design. Cycle stability calculations are

performed by SPC. Therefore, the probability of thermal hydraulic

instability is not significantly increased.

Evaluations of the Dresden Emergency Procedures and UFSAR

Chapter 15 AOOs are being performed to ensure that the use of the

ATRIUM-9B fuel at Dresden does not alter any assumptions previously

made in evaluating the radiological consequences of an accident at

Dresden Units 2 and 3. Therefore, the radiological consequences of

accidents are not significantly increased.

Methods approved by the NRC are being used in the evaluation of

fuel performance during normal and abnormal operating conditions.

The ComEd and SPC methods to be used for the cycle specific

transient analyses have been previously NRC approved. The proposed

methodologies are administrative in nature and do not significantly

affect any accident precursors or accident results; as such, the

proposed change to the listing of the SPC methodologies for Dresden

does not significantly increase the probability or consequences of

any previously evaluated accidents.

The description of the fuel is modified to include the water box

design of the NRC approved ATRIUM-9B fuel type.

Review of the above concludes that the probability of occurrence

and the consequences of an accident previously evaluated in the

safety analysis report have not been significantly increased.

* * * * *

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

from any accident previously evaluated:

Creation of the possibility of a new or different kind of

accident would require the creation of one or more new precursors of

that accident. New accident precursors may be created by

modifications of the plant configuration, including changes in

allowable modes of operation.

Licensing Methods and Models

The proposed Technical Specification amendment reflects

previously approved SPC methodology used to analyze normal

operations, including AOOs, and to determine the potential

consequences of accidents. In accordance with FSAR Chapter 15, the

same accidents and transients will be analyzed with the new fuel and

method as have been previously performed. As stated above, the

proposed changes do not permit modes of reactor operation which

differ from those currently permitted; therefore, the possibility of

a new or different kind of accident is not created. Plant support

equipment is not affected by the proposed changes; therefore, no new

failure modes are created.

New Fuel Design

The basic design concept of a 9x9 fuel pin array with an

internal water box has been used in various lead assembly programs

and in reload quantities in Europe since 1986. WNP-2 has loaded

reload quantities since 1991. Eight lead ATRIUM-9B assemblies were

loaded into Dresden 2 during Cycle 15. Approximately 650 water box

assemblies have been irradiated in the United States through 1995,

with a substantially higher number being irradiated overseas. The

NRC has reviewed and approved the ATRIUM-9B fuel design (Reference:

ANF-89-014(P)(A) Revision 1 Supplement 1, Generic Mechanical Design

for Advanced Nuclear Fuels 9X9-IX and 9X9-9X BWR Reload Fuel). The

similarities in fuel design and operation between the ATRIUM-9B and

the 9x9-2, and the previous Boiling Water Reactor experience with

Siemens fuel, indicate there would be no new or different types of

accidents for Dresden than have been considered for the existing

fuel. Therefore, the use of ATRIUM-9B fuel at Dresden does not

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

any accident previously evaluated.

* * * * *

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

the following reasons:

The existing margin to safety is provided by the existing

acceptance criteria (e.g., 10 CFR 50.46 limits). The proposed

Technical Specification amendment reflects previously approved SPC

methodology used to demonstrate that the existing acceptance

criteria are satisfied. The revised LOCA methodology has been

previously reviewed and approved by the USNRC for application to

reload cores of BWRs. References for the Licensing Topical Reports

which document this methodology, and include the Safety Evaluation

Reports prepared by the USNRC, are added to the Reference section of

the Technical Specifications as part of this amendment.

Licensing Methods and Models

The proposed amendment does not involve changes to the existing

operability criteria. NRC approved methods and established limits

(implemented in the COLR) ensure acceptable margin is maintained.

The ComEd and SPC reload methodologies for the ATRIUM-9B reload

design are consistent with the Technical Specification Bases. The

Limiting Conditions for Operation are taken into consideration while

performing the cycle specific and generic reload safety analyses.

USNRC approved methods are listed in Specification 6.9.A of the

Technical Specifications.

Analyses performed with USNRC-approved methodology have

demonstrated that fuel design and licensing criteria will be met

during normal and abnormal operating conditions. The same margins of

safety will continue to be utilized by SPC (e.g., limits on peak

cladding temperature, cladding oxidation, plastic strain).

Therefore, there is not a significant reduction in the margin of

safety.

New Fuel Design

The exterior dimensions of the ATRIUM-9B fuel assembly result in

equivalent clearances relative to the co-resident 9x9-2 fuel. Thus,

no increased interactions with the adjacent control blade and

nuclear instrumentation are created. The change does not adversely

impact equipment important to safety; 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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Morris Area Public Library

District, 604 Liberty Street, Morris, Illinois 60450.

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

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

NRC Project Director: Robert A. Capra.

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

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

Date of amendment request: March 18, 1997.

Description of amendment request: The proposed amendments would

change the Technical Specifications (TS) by increasing the High

Pressure Coolant Injection (HPCI) isolation setpoint from greater than/

equal to 80 psig to greater than/equal to 100 psig. The licensee has

requested the change to ensure consistency between the Updated Final

Safety Analysis Report (UFSAR), design basis documents and the TS. The

function of the setpoint is to assure the HPCI turbine steam supply is

isolated in the event that the reactor scram supply pressure falls

below the stall pressure of the HPCI turbine and the system seals are

no longer effective in controlling the release of potentially

contaminated steam.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated because of the

following:

The Low Reactor Pressure isolation of the HPCI steam supply

lines is provided to prevent damage to the HPCI turbine when the

reactor steam pressure has decreased below that required to provide

adequate motive force to operate the system. The steam supply

isolation low reactor pressure setpoint is not an assumed initiator

or contributor to any previously evaluated accident and therefore

this change does not involve an increase in the probability of an

accident previously evaluated at Dresden Station.

The Lower Reactor Pressure isolation of the HPCI steam supply

lines is described in the

[[Page 17229]]

plant safety analysis as a backup protection to other system and

facility design features which provide assurance that accident

transients will not result in failures of the system which

contribute significantly to the consequences of the initiating

accident. The low reactor pressure isolation signal provides backup

to other isolation signals to ensure isolation will occur,

minimizing the radiation dose as a result of steam leakage past the

turbine seals in the event of a locked rotor due to damage from

liquid carryover due to postulated swell in the reactor vessel.

These analyses assume the isolation function occurs at 100 psig,

and the proposed setpoint of greater than or equal to 100 psig is

consistent and conservative with respect to these assumptions.

Because the isolation function is not an accident initiator and the

revised setpoint ensures that the isolation function continues to

minimize radiological consequences, the consequences of any accident

previously evaluated is not increased by the proposed changes.

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

from any accident previously evaluated because:

The proposed change administratively increases the Low Reactor

Vessel Pressure trip setpoint which initiates HPCI isolation. This

change does not result in any new or different modes of operation.

The proposed change increases the setpoint at which the HPCI turbine

steam supply will be isolated as the reactor vessel pressure

decreases following a postulated accident. The proposed new setpoint

is conservative with respect to the existing TS limit, i.e. the new

limit of greater than or equal to 100 psig is consistent and

permitted by the existing limit of greater than or equal to 80 psig.

The change assures that the Trip Setpoint in the TS accurately

reflects the design basis and UFSAR described limits.

Because the proposed change does not result in any new modes of

plant operation and administratively increases the system isolation

setpoint in a conservative manner, the proposed change does not

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

those previously evaluated.

3. Involve a significant reduction in the margin of safety

because:

The Trip Setpoint provides assurance that the HPCI turbine

cannot be operated with a steam supply pressure too low to drive the

turbine and pump. The isolation assures that the turbine does not

stall and minimizes the potential for the release of radioactivity

which results from steam leakage past the turbine seals. The

proposed change increases the setpoint, ensuring that the required

isolation occurs at a higher pressure which is more conservative,

i.e. it assures the turbine is isolated before the inlet steam

pressure falls to the stall pressure of the HPCI turbine and leakage

occurs. The greater than or equal to 100 psig limit is well below

the range of reactor vessel pressure for which HPCI is required to

perform its safety function. Therefore, the margin of safety

provided by the function of the HPCI isolation on low reactor vessel

pressure is increased by the proposed TS change, and this change

will not involve a reduction in the margin of safety.

As described, the proposed amendment for Dresden will not reduce

the availability of systems required to mitigate accident

conditions. Neither are new or significantly different modes of

operation proposed. Therefore, the proposed change does not involve

a significant reduction in the margin of safety.

Guidance has been provided in ``Final Procedures and Standards

on No Significant Hazards Considerations,'' Final Rule, 51 FR 7744,

for the application of standards to license change requests for

determination of the existence of significant hazards

considerations. This document provides examples of amendments which

are and are not considered likely to involve significant hazards

considerations.

This proposed amendment does not involve any irreversible

changes, a significant relaxation of the criteria used to establish

safety limits, a significant relaxation of the bases for the

limiting safety system settings or a significant relaxation of the

bases for the limiting conditions for operations. Therefore, based

on the guidance provided in the Federal Register and the criteria

established in 10 CFR 50.92(c), the proposed change does not

constitute 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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Morris Area Public Library

District, 604 Liberty Street, Morris, Illinois 60450.

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

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

NRC Project Director: Robert A. Capra.

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

County, Michigan

Date of amendment request: December 27, 1995, as supplemented

September 4, October 18, and November 26, 1996.

Description of amendment request: The proposed amendment would

revise technical specifications (TS) related to electrical power

systems. The proposed changes include revisions to limiting conditions

for operation (LCO), LCO applicability and action statements, allowed

outage times (AOT), surveillance requirements (SR), and administrative

controls. The changes add new requirements, revise or delete existing

requirements, relocate certain existing requirements to other licensee

controlled documents, and editorially restructure the proposed

requirements to closely emulate the electrical power system

requirements of NUREG-1432, ``Standard Technical Specifications for

Combustion Engineering Plants,'' (STS). The proposed requirements

differ from the requirements of the STS where necessary to reflect

features unique to the Palisades design. Each proposed change has been

classified by the licensee as Administrative, Relocated, More

Restrictive, or Less Restrictive.

Basis for proposed no significant hazards consideration

determination: A proposed amendment to an operating license for a

facility involves no significant hazards consideration if operation of

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

Involve a significant increase in the probability or consequences of an

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

different kind of accident from any previously evaluated; or (3)

involve a significant reduction in a margin of safety. 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:

Evaluation of ADMINISTRATIVE, RELOCATED, and MORE RESTRICTIVE

changes:

ADMINISTRATIVE changes and RELOCATED changes move requirements,

either within the TS or to documents controlled under 10 CFR 50.59,

or [clarify] existing TS requirements, without affecting their

technical content. Since ADMINISTRATIVE and RELOCATED changes do not

alter the technical content of any requirements, they cannot involve

a significant increase in the probability or consequences of an

accident previously evaluated, create the possibility of a new or

different kind of accident from any previously evaluated, or involve

a significant reduction in a margin of safety.

MORE RESTRICTIVE changes only add new requirements, or revise

existing requirements to result in additional operational

restrictions. Since the TS, with all MORE RESTRICTIVE changes

incorporated, will still contain all of the requirements which

existed prior to the changes; MORE RESTRICTIVE changes cannot

involve a significant increase in the probability or consequences of

an accident previously evaluated, create the possibility of a new or

different kind of accident from any previously evaluated, or involve

a significant reduction in a margin of safety.

Evaluation of LESS RESTRICTIVE changes:

1. Do these LESS RESTRICTIVE changes involve a significant

increase in the probability or consequences of an accident

previously evaluated?

Change 3 revised the requirement for operable AC sources, using

more general wording than the existing TS. The existing LCO requires

that two explicitly specified transformers be operable; the proposed

LCO requires that two qualified offsite circuits be operable. The

proposed LCO will allow

[[Page 17230]]

substitution of Safeguards Transformer 1-1 for Station Power

Transformer 1-2 as a required AC source, but the quantity and

quality of required offsite AC sources is unaffected. Since the

capability and qualification of Safeguards Transformer 1-1 are

equivalent to those of the Station Power transformer, neither the

probability or consequences of an accident previously evaluated will

be increased.

Change 10 is less restrictive only in its allowance of a 72 hour

AOT for an inoperable offsite source instead of the 24 hour AOT

currently required. The change also makes a considerably more

restrictive change by eliminating the allowance, based on submittal

of a report, for continuous operation with Startup Transformer 1-2

inoperable. Changing an AOT, alone, cannot increase the probability

or consequences of an accident previously evaluated.

Change 14 allows, for an inoperable DG [diesel generator],

verification that no common cause failure is involved in lieu of

test starting the other DG. The intent of the test starting

requirement is to verify that there is no common cause failure which

also makes the other DG inoperable. The proposed action statement

thereby accomplishes the same objective as that it replaces. Since

the proposed action statement accomplishes the same objective as the

one it replaces, operation in accordance with the proposed change

will not increase the probability or consequences of an accident

previously evaluated.

Change 21 revises the SR for the DG starting test. [The ``Less

Restrictive'' elements of the change eliminate the requirement to

vary use of the A and B starting circuits for each monthly test,

because the DG is not assumed to be single failure proof; and

eliminate requirements that the DGs be manually started and that

they be synchronized from the control room, because no practical

alternatives exist for accomplishing these actions]. The proposed

change does not alter any plant operating conditions, operating

practices, equipment settings, or equipment capabilities. Therefore,

operation of the facility in accordance with the proposed change

will not involve an increase in the probability of an accident.

Change 21 requires more rigorous testing of the DGs than required by

the existing Technical Specifications. The more rigorous testing is

intended to provide additional assurance that the DGs are capable of

performing their design function and should, therefore, involve a

reduction, rather than an increase, in the consequences of those

accidents previously evaluated.

Change 25 revises the SR for testing the fuel transfer system.

The proposed change does not alter any plant operating conditions,

operating practices, equipment settings, or equipment capabilities.

Therefore, operation of the facility in accordance with the proposed

change will not involve an increase in the probability of an

accident. The only ``Less Restrictive'' feature of proposed SR is

test interval extension from ``each month'' to ``each 92 days.''

Changing a surveillance frequency, alone, cannot increase the

probability or consequences of an accident previously evaluated.

Change 26 revises the station battery SRs. The proposed monthly

and quarterly battery SRs contain all of the test requirements of

the existing SRs with two exceptions: (1) The proposed interval for

measuring each cell voltage is ``each 92 days'' instead of the

existing ``every month'' and (2) the requirement to record the

amount of water added has been deleted. Changing a surveillance

frequency or deleting a maintenance record cannot increase the

probability or consequences of an accident previously evaluated.

2. Do changes create the possibility of a new or different kind

of accident from any previously evaluated?

Change 3 only involves the specified offsite power sources.

Since the Loss of Offsite Power is already considered in the

accident analyses, operating the facility in accordance with Change

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

accident from any previously evaluated.

Change 10 revises an AOT; Change 14 revises a required action;

Change 21 revises a testing requirement; Changes 25 and 26 revise a

surveillance interval; and Change 26 deletes the requirement for a

maintenance record. None of these proposed changes alter any plant

operating conditions, operating practices, equipment settings, or

equipment capabilities. Therefore, operation of the facility in

accordance with the proposed changes will not create the possibility

of a new or different kind of accident from any previously

evaluated.

3. Do changes involve a significant reduction in a margin of

safety?

Change 3 does not alter the quantity or quality of offsite

sources required to be available. Therefore, operating the facility

in accordance with the proposed change will not involve a reduction

in a margin of safety.

Change 10 revises an AOT; Change 14 revises a required action,

Change 21 revises a testing requirement; Changes 25 and 26 revise a

surveillance interval; and Change 26 deletes the requirement for a

maintenance record. These proposed changes do not alter any plant

operating conditions, operating practices, equipment settings, or

equipment capabilities. Therefore, operating the facility in

accordance with the proposed change will not involve a reduction in

a margin of safety.

The licensee's September 4, 1996, supplement stated that three of

the proposed changes contained in the supplement were not addressed in

the December 27, 1995, no significant hazards analysis. The changes

involved TS requirements that would be deleted. Equivalent requirements

would be incorporated in the FSAR or other documents subject to the

controls of 10 CFR 50.59. The licensee's analysis of the issue of no

significant hazards consideration for these changes is presented below:

1. Do changes which relocate a requirement from the TS to

documents which are controlled under 10 CFR 50.59 involve a

significant increase in the probability or consequences of an

accident previously evaluated?

10 CFR 50.59 specifically prohibits [without obtaining prior NRC

review and approval] changes to the facility as described in the

safety analysis report, and to procedures described in the safety

analysis report ``if the probability of occurrence or the

consequences of an accident or malfunction of equipment important to

safety previously evaluated in the safety analysis report may be

increased''. Since the conditions which limit changes performed

under 50.59 are more restrictive than the conditions which define

changes considered to involve a significant hazards consideration,

relocation of a requirement from the TS to the FSAR [Final Safety

Analysis Report] or to documents which are referenced by the FSAR

cannot involve a significant increase in the probability or

consequences of an accident previously evaluated.

2. Do changes which relocate a requirement from the TS to

documents which are controlled under 10 CFR 50.59 create the

possibility of a new or different kind of accident from any

previously evaluated?

10 CFR 50.59 specifically prohibits [without obtaining prior NRC

review and approval] changes to the facility as described in the

safety analysis report, and to procedures described in the safety

analysis report ``if a possibility for an accident or malfunction of

a different type than any evaluated previously in the safety

analysis report may be created''. Since the conditions which limit

changes performed under 50.59 are more restrictive than the

conditions which define changes considered to involve a significant

hazards consideration, relocation of a requirement from the TS to

the FSAR or to documents which are referenced by the FSAR cannot

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

any previously evaluated.

3. Do these changes which relocate a requirement from the TS to

documents which are controlled under 10 CFR 50.59 involve a

significant reduction in a margin of safety?

10 CFR 50.59 specifically prohibits [without obtaining prior NRC

review and approval] changes to the facility as described in the

safety analysis report, and to procedures described in the safety

analysis report ``if the margin of safety as defined in the basis

for any technical specification is reduced''. Since the conditions

which limit changes performed under 50.59 are more restrictive than

the conditions which define changes considered to involve a

significant hazards consideration, relocation of a requirement from

the TS to the FSAR or to documents which are referenced by the FSAR

cannot involve a significant reduction in a margin of safety.

The NRC staff has reviewed the licensee's analyses 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.

[[Page 17231]]

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

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

NRC Project Director: John N. Hannon.

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

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

Pennsylvania

Date of amendment request: March 10, 1997.

Description of amendment request: The proposed amendment would

modify Unit 1 Technical Specification (TS) 5.2.1 to add ZIRLO as fuel

assembly material and add reference to Nuclear Regulatory Commission

approved Topical Report, WCAP-12610, ``Vantage+ Fuel Assembly Reference

Core Report'', to TS 6.9.1.12 for both units.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The methodologies used in the accident analyses have been

modified to reflect the requirements provided in WCAP-12610,

VANTAGE+ Fuel Assembly Reference Core Report. Reference to this NRC

approved ZIRLO topical report has been added to Specification

6.9.1.12, for both units to ensure the analytical methods used to

determine the core operating limits are consistent with those

previously approved by the NRC. The proposed changes do not change

or alter the design assumptions for the systems or components used

to mitigate the consequences of an accident. Use of ZIRLO fuel rod

material does not adversely affect fuel performance or impact

nuclear design methodology. Therefore, accident analysis results are

not impacted.

The operating limits will not be changed and the analysis

methods to demonstrate operation within the limits will remain in

accordance with NRC approved methodologies. Other than the changes

to the fuel assemblies, there are no physical changes to the plant

associated with this technical specification change. A safety

analysis will continue to be performed for each cycle to demonstrate

compliance with all fuel safety design bases.

VANTAGE 5H fuel assemblies with ZIRLO fuel rods meet the same

fuel assembly and fuel rod design bases as other VANTAGE 5H fuel

assemblies. In addition, the 10 CFR 50.46 criteria are applied to

the ZIRLO fuel rods. The use of these fuel assemblies will not

result in a change to the reload design and safety analysis limits.

Since the original design criteria are met, the ZIRLO fuel rods will

not be an initiator for any new accident. The fuel rod material is

similar in chemical composition and has similar physical and

mechanical properties as Zircaloy-4. Thus, the fuel rod integrity is

maintained and the structural integrity of the fuel assembly is not

affected. ZIRLO improves corrosion performance and dimensional

stability. No concerns have been identified with respect to the use

of an assembly containing a combination of Zircaloy-4 and ZIRLO fuel

rods.

The dose predictions in the safety analyses are not sensitive to

the fuel rod material used; therefore, the radiological consequences

of accidents previously evaluated in the safety analysis remain

valid. A reload analysis is completed for each cycle, in accordance

with NRC approved methodologies. Therefore, the proposed change does

not involve a significant increase in the probability or

consequences of an accident previously evaluated.

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

kind of accident from any accident previously evaluated?

VANTAGE 5H fuel assemblies with ZIRLO fuel rods satisfy the same

design bases as those used for other VANTAGE 5H fuel assemblies. All

design and performance criteria continue to be met and no new

failure mechanisms have been identified. The ZIRLO fuel rod material

offers improved corrosion resistance and structural integrity.

The proposed changes do not affect the design or operation of

any system or component in the plant. The safety functions of the

related structures, systems, or components are not changed in any

manner, nor is the reliability of any structure, system, or

component reduced. The changes do not affect the manner by which the

facility is operated and do not change any facility design feature,

structure, or system. No new or different type of equipment will be

installed. Since there is no change to the facility or operating

procedures, and the safety functions and reliability of structures,

systems, or components are not affected, the proposed changes do not

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

any accident previously evaluated.

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

of safety?

The use of Zircaloy-4, ZIRLO, or stainless steel filler rods in

fuel assemblies will not involve a significant reduction in the

margin of safety because analyses using NRC approved methodology

will be performed for each configuration to demonstrate continued

operation within the limits that assure acceptable plant response to

accidents and transients. These analyses will be performed using NRC

approved methods that have been approved for application to the fuel

configuration.

Use of ZIRLO as fuel rod material does not change the VANTAGE 5H

reload design and safety analysis limits. The use of these fuel

assemblies will take into consideration the normal core operating

conditions allowed in the technical specifications. For each reload

core, the fuel assemblies will be evaluated using NRC approved

reload design methods, including consideration of the core physics

analysis peaking factors and core average linear heat rate effects.

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

amendment request does not result in a significant reduction in

margin with respect to plant safety as defined in the UFSAR [Updated

Final Safety Analysis Report] or any plant 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 request involves no significant hazards consideration.

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

663 Franklin Avenue, Aliquippa, PA 15001.

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

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

NRC Project Director: John F. Stolz.

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

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

Date of amendment request: July 17, 1996.

Description of amendment request: The proposed amendment would

reflect that the name of Louisiana Power & Light Company, which is

licensed to own and possess Waterford 3, has been changed to Entergy

Louisiana, Inc.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Will operation of the facility in accordance with this

proposed change involve a significant increase in the probability or

consequences of an accident previously evaluated?

Response: No

The proposed change documents changing the legal name of the

company. The proposed change will not affect any other obligations.

The company will still own all of the same assets, they still serve

the same customers, and all existing obligations and commitments

will continue to be honored. Therefore, the proposed change will not

involve a significant increase in the probability or consequences of

any accident previously evaluated.

2. Will operation of the facility in accordance with this

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

accident from any accident previously evaluated?

Response: No

The administrative changes in the Operating License requirements

do not involve any change in the design of the plant. Therefore, the

proposed change will not create the possibility of a new or

different

[[Page 17232]]

kind of accident from any accident previously evaluated.

3. Will operation of the facility in accordance with this

proposed change involve a significant reduction in a margin of

safety?

Response: No

The proposed change is administrative in nature and does not

reduce the level of safety imposed by any current requirement.

Therefore, the proposed change will not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: University of New Orleans

Library, Louisiana Collection, Lakefront, New Orleans, LA 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: October 16, 1996.

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) action requirements 3.2.1 and 3.2.4

and their associated surveillance requirements to extend the allowable

time for the Core Operation Limit Supervisory System (COLSS) to be out

of service by monitoring for adverse trends in the linear heat rate

(LHR) and departure from nucleate boiling (DNBR) limits.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Will operation of the facility in accordance with this

proposed change involve a significant increase in the probability or

consequences of an accident previously evaluated?

Response: No

The proposed change does not modify the requirement to operate

within the alternate LHR and DNBR limits nor does it modify the

actual LHR or DNBR limits themselves. In the case of exceeding a

COLSS calculated [power operating limit] POL, Entergy agrees that

corrective action should be initiated promptly to bring the LHR and

DNBR within their respective limits and, in this case, a 15 minute

time limit is appropriate. However, in the case of exceeding a [core

protection calculator] CPC calculated operating limit following the

loss of COLSS, it is clear that simply because COLSS execution was

lost does not mean that the plant is operating outside the range of

conditions assumed in the Chapter 15 Safety Analysis and, in this

case, a 15 minute time limit is not appropriate. An increase from 2

hours to 8 hours to regain the monitoring capabilities of COLSS

would not significantly increase the probability of exceeding the

actual LHR or DNBR power operating limits since the increase in

COLSS out-of-service time will be compensated for by monitoring for

adverse trends of the important CPC calculated parameters (DNBR

Margin and LHR). Further, since the proposed change will result in

maintaining steady-state conditions while monitoring for adverse

trends, it will be easier for the operators to detect any abnormal

occurrence that has the potential to degrade either the LHR or the

DNBR.

The primary consideration in extending the COLSS out of service

time limit is the remote possibility of a slow, undetectable

transient that degrades the LHR and/or DNBR slowly over the 8 hour

period and is then followed by an [anticipated operational

occurrence] AOO or an accident. The parameters normally monitored by

COLSS which have the potential for degrading the LHR and DNBR if no

corrective action is taken are: Reactor Coolant System (RCS) flow

rate, axial and radial power distributions, core inlet temperature,

core power, RCS pressure and azimuthal tilt. Of these parameters,

core inlet temperature, core power, and RCS pressure are easily

monitored by the plant operators using various safety-grade,

Redundant Control Room indications and, therefore, changes in these

parameters are readily apparent. Further, operating experience at

Waterford 3 and other [Combustion Engineering] CE nuclear steam

supply systems using the same reactor coolant pumps (RCPs) as

Waterford has shown that measurable changes in RCP Ps

(which COLSS uses to calculate RCS flow) are very rare and when they

do occur, involve abrupt step changes in flow which are readily

apparent; hence, the probability of a slow degradation in the RCS

flow rate is exceedingly small. Thus, the parameters that

comparatively (although still remote) pose the most potential for a

degradation in the core thermal margin when COLSS is out of service

relate to the axial and radial core power distributions and the

azimuthal tilt. These parameters are discussed below.

Axial xenon oscillations are a normal consequence of the

Waterford 3 core design, particularly near the end of core life. As

a result, Waterford 3 operations personnel are instructed, per

operating procedure OP-10-001, General Plant Operations, to maintain

strict control over the axial power shape in the core. Although the

primary reason for axial shape control is to maintain an even fuel

burnup throughout the core, it also results in maintaining the axial

power shapes well within the limits assumed in the safety analysis.

Typically, axial shape control practiced at Waterford 3 maintains

the axial shape index (ASI) within 0.05 ASI units of the equilibrium

shape index (ESI), which is normally very near 0.0.

Hypothetically, the most severe situation which could be

postulated to occur, although again remote, would be if COLSS

execution was lost just when the plant operators were ready to take

manual action to return the ASI value to within the ESI + 0.05

control band. Since a full xenon oscillation takes approximately 26

hours, there would be about 6 hours from the time that control

action would normally be taken to the time that the ASI reached its

peak value (i.e., it takes one quarter cycle for the ASI to travel

from its ESI value to its peak value). Since abnormal operating

procedure OP-901-501, PMC or Core Operating Limit Supervisory System

Inoperable, will be revised to require the CPC calculated LHR and

DNBR trends to be monitored every 15 minutes (see below), any

significant change in the axial shape index will be apparent through

a change in these CPC calculated values. Hence, due to the attention

given the axial power distribution, both when COLSS is in service as

well as when COLSS is out of service it is very improbable that a

change in ASI during eight hours of steady-state operation with

COLSS out of service could be either undetected or lead to a

condition that placed the reactor outside the range of initial

conditions that were assumed in the safety analysis.

With regards to azimuthal tilt, there is very rarely any

significant change in this parameter as long as all [control element

assembly] CEAs are properly aligned. The only real contributor to a

rapid increase in azimuthal tilt would be an inadvertent CEA drop;

however, since the probability of a CEA drop is very low, the

likelihood of this event occurring within the eight hour time limit

is even lower. In the unlikely event that a CEA drop did occur, the

Control Element Assembly Calculators (CEACs) provide a safety-grade,

redundant means of alerting the operators that corrective action is

necessary. Thus, the potential for a degradation in azimuthal tilt

during eight hours of steady-state operation following the loss of

COLSS is both highly unlikely and relatively easy to detect using

instrumentation already available in the Control Room.

As previously stated, upon approval of the proposed change plant

personnel will revise abnormal operating OP-901-501, PMC or Core

Operating Limit Supervisory System Inoperable, to monitor for

adverse trends of the CPC calculated values of LHR and DNBR.

Currently, this procedure requires that the monitoring frequency for

LHR and DNBR be increased to once every 15 minutes on a loss of

COLSS.

Extending the time to restore the CPC calculated LHR and DNBR to

within the acceptable operating range from 2 hours to 8 hours is

being proposed to assure that COLSS can be restored thus decreasing

the probability of an avoidable challenge to the reactor protection

system (RPS) during a power reduction. It is possible that the

required power reductions may exceed 25% near the end of the fuel

cycle. These large power reductions result in a rapid increase in

xenon concentration, changes in ASI, and a subsequent decrease in

cold leg temperature (T-cold) that may be difficult to control.

Accordingly, given the potential for

[[Page 17233]]

power reductions of this magnitude, it is appropriate to extend the

time allowed to restore COLSS so that a power reduction may be

unnecessary.

Taken in total, the proposed changes will reduce the number of

potentially unnecessary power reductions by allowing more time for

COLSS to be restored along with the advantages of trend monitoring

in detecting an adverse trend expeditiously. The proposed change

will result in significant operational benefits while continuing to

maintain a high degree of confidence that the core conditions remain

well within the range of values assumed in the safety analysis.

Therefore, the proposed change will not involve a significant

increase in the probability or consequences of any accident

previously evaluated.

2. Will operation of the facility in accordance with this

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

accident from any accident previously evaluated?

Response: No

The proposed change does not alter the current power operating

limits nor does it involve any changes to COLSS or CPC software.

There has been no physical change to plant systems, structures or

components nor will the proposed change affect the ability of any of

the safety-related equipment required to mitigate AOOs or accidents.

The only significant change associated with the proposed amendment

involves changes to the operating procedures used when COLSS is out-

of-service. All revisions to operating procedures will be reviewed

and approved by appropriate plant personnel as required by the

Administrative Controls (Section 6) in the Waterford 3 Technical

Specifications. Therefore, the proposed change will not create the

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

previously evaluated.

3. Will operation of the facility in accordance with this

proposed change involve a significant reduction in a margin of

safety?

Response: No

The intent of [limiting conditions for operation] LCOs 3.2.1 and

3.2.4 is to maintain the reactor within the range of initial

conditions that was assumed in the Safety Analysis. Maintaining the

LHR within the specified range ensures that in the event of a LOCA,

the fuel cladding temperature will not exceed the 2200 deg.F limit

imposed by 10CFR46 [10 CFR Part 46]. Maintaining the DNBR within the

specified range ensures that no AOO will result in a violation of

the [Specified Acceptable Fuel Design Limits] SAFDLs and that no

postulated accident will result in consequences more severe than

those described in Chapter 15 of the [Final Safety Analysis Report]

FSAR. Since there has been no change to the requirement to operate

the reactor within the LHR and DNBR limits and no change to the

actual LHR and DNBR limits themselves, the accident analyses

described in Chapter 15 of the FSAR will not be affected and will

therefore remain bounding.

The proposed change will reduce the number of potentially

unnecessary power reductions along with the rate at which the power

reductions are accomplished. Maintaining steady-state conditions for

up to eight hours after the loss of COLSS while monitoring the CPC

LHR/DNBR for trends, provides plant personnel with a reasonable

period of time to return COLSS to service while continuing to

maintain a high degree of confidence that the core conditions remain

well within the range of values assumed in the safety analysis. In

fact, monitoring for trends in LHR and DNBR Margin increases the

margin of safety by allowing the anticipation of degradation in LHR

or DNBR Margin. Moreover, by reducing the number of plant transients

there will be an attendant reduction in probability of an AOO and

subsequent RPS actuation. Therefore, the proposed change will not

involve a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: University of New Orleans

Library, Louisiana Collection, Lakefront, New Orleans, LA 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: October 16, 1996.

Description of amendment request: The following changes to the

Waterford Steam Electric Station, Unit 3, Technical Specifications are

proposed: 1) Relocation of certain administrative controls to the

Quality Assurance Program Manual (QAPM) as described in Nuclear

Regulatory Commission Administrative Letter 95-06, ``Relocation of

Technical Administrative Controls related to Quality Assurance''; 2)

Change shift coverage from 8-hour day, 40-hour weeks to an option of 8

or 12 hour days and nominal 40-hour weeks; 3) Make certain editorial

changes to the titles of certain organizational positions.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Will operation of the facility in accordance with this

proposed change involve a significant increase in the probability or

consequences of an accident previously evaluated?

Response: No

The conditions as they exist in the present Technical

Specifications do not have an affect on either the probability or

consequences of a previously evaluated accident. These changes also

will have no impact to increase either the probability or

consequences of a previously evaluated accident.

The proposed changes will have no affect on design basis

accidents nor will the change directly affect any material condition

of the plant that could directly contribute to causing or mitigating

the effects of an accident.

Therefore, the proposed changes will not involve a significant

increase in the probability or consequences of any accident

previously evaluated.

2. Will operation of the facility in accordance with this

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

accident from any accident previously evaluated?

Response: No

The proposed changes will not alter the operation of the plant

or the manner in which it is operated. The changes do not involve a

design change and do not introduce any new failure modes.

Therefore, the proposed changes will not create the possibility

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

evaluated.

3. Will operation of the facility in accordance with this

proposed change involve a significant reduction in margin of safety?

Response: No

The proposed changes are administrative in nature and affect

only Section 6.0 of the Technical Specifications. The Waterford 3

margins of safety are defined in Sections 2 through 5 and are

unaffected by these changes. Moving the reviews from the TS to the

QAPM will have no affect on the margin of safety because reviews

will still be performed. The only difference is the reviews will be

administratively controlled by the QAPM. The QAPM is controlled by

10CFR50.54 so no changes can be made which would lessen these

commitments (i.e., remove or reduce the requirement for procedure

reviews) without prior NRC approval.

Changing from an 8 hour to an 8 or 12 hour shift will not have

an adverse impact on personnel performance. The NRC study documented

in NUREG CR-4248 has identified that personnel errors have decreased

and productivity has increased where this change has been

implemented.

Therefore, the proposed changes will not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: University of New Orleans

[[Page 17234]]

Library, Louisiana Collection, Lakefront, New Orleans, LA 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: March 27, 1997.

Description of amendment request: The proposed change modifies

Technical Specification 3/4.5.2, ``ECCS Subsystems Modes 1, 2, and 3.''

The proposed change adds a surveillance requirement to verify the

Emergency Core Cooling System (ECCS) piping is full of water at least

once per 31 days. A change to the Technical Specification Basis 3/4.5.2

and 3/4.5.3 has been included to support this change.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Will operation of the facility in accordance with this

proposed change involve a significant increase in the probability or

consequences of an accident previously evaluated?

Response: No.

The proposed change will not affect the assumptions, design

parameters, or results of any accident previously evaluated. The

proposed change does not add or modify any existing equipment. The

proposed change adds a new surveillance requirement which will

minimize the likelihood of a pressure transient occurring during

system startup and provide increased assurance that the ECCS will

perform its design basis function when needed. The new [low pressure

safety injection] LPSI and [high pressure safety injection] HPSI

vent valves which may be manipulated during this surveillance will

be administratively controlled and will be locked close when not in

use to prevent the possibility of a flow diversion. This

surveillance requirement is consistent with NUREG 1432.

Therefore, the proposed change will not involve a significant

increase in the probability or consequences of any accident

previously evaluated.

2. Will operation of the facility in accordance with this

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

accident from any accident previously evaluated?

Response: No.

While new vent lines are being installed under 10CFR50.59, this

proposed change adds only a new surveillance requirement to

Technical Specification 3/4.5.2 and therefore does not involve

modifications to any existing equipment. The new vent valves, when

required, will be operated and controlled in the same manner as

existing LPSI and HPSI vent valves. The new LPSI and HPSI vent

valves will be administratively controlled and will be locked close

when not in use. This surveillance requirement is consistent with

NUREG 1432.

Therefore, the proposed change will not create the possibility

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

evaluated.

3. Will operation of the facility in accordance with this

proposed change involve a significant reduction in a margin of

safety?

Response: No.

The functionality of ECCS is maintained such that it is capable

of performing its design function as assumed in the Updated Final

Safety Analysis Report. Verifying the ECCS is full of water at least

once per 31 days will minimize the likelihood of a pressure

transient occurring during system startup and provide increased

assurance that the ECCS will perform its design basis function when

needed. This surveillance requirement is consistent with NUREG 1432.

Therefore, the proposed change will not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: University of New Orleans

Library, Louisiana Collection, Lakefront, New Orleans, LA 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: March 27, 1997.

Description of amendment request: The proposed change modifies

Technical Specification (TS) surveillance requirements 4.5.2.d.3 and

4.5.2.d.4. The proposed change specifies granular trisodium phosphate

dodecahydrate (TSP), increases the minimum required amount of TSP that

is maintained in containment during power operation, and adjusts the

TSP sampling requirement accordingly. A change to the TS Basis 3/4.5.2

has been included to support this change.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Will operation of the facility in accordance with this

proposed change involve a significant increase in the probability or

consequences of an accident previously evaluated?

Response: No.

Granular trisodium phosphate dodecahydrate is stored in the

containment lower level to raise the pH of the sump and spray water

following a LOCA. As the pH of the water increases, more radioactive

iodine is kept in solution and the amount of airborne radioactive

leakage is decreased. This also lessens the potential for boric acid

solution reacting with galvanized metal in containment to release

hydrogen. An additional advantage of a higher pH is the beneficial

reduction in chloride stress corrosion cracking of metal components

in the containment following an accident.

This chemical is an accident mitigator, not an accident

initiator in that it is not used until after an accident has

occurred. At the time it goes into solution, the accident has

occurred, containment spray has been activated and water has

collected in the sump. Therefore, increasing the Technical

Specification minimum amount verified to be in containment or

changing the sample solution and sample size will not involve a

significant increase in the probability of an accident previously

evaluated.

At the time TSP goes into solution, the accident has occurred,

containment spray has been activated and water has collected in the

containment sump. At Waterford 3, the iodine partition factor is a

constant 50% and does not vary with pH as allowed in the Standard

Review Plan (SRP) revision 1. The curve in SRP 6.5.2 revision 1

allows a partition factor of at least 50% for containment water at a

pH of 6.5 or less. The partition factor increases as pH rises. But,

the curve is based on sodium hydroxide which is much more reactive

than TSP. Therefore, increasing the Technical Specification minimum

amount verified to be in the containment, and corresponding sample

size, will not involve any significant increase in the consequences

probability of an accident because no credit is taken for reducing

the amount of volatized iodine normally associated with a 7.0 pH

solution.

Therefore, the proposed change will not involve a significant

increase in the probability or consequences of any accident

previously evaluated.

2. Will operation of the facility in accordance with this

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

accident from any accident previously evaluated?

Response: No.

The addition of more TSP does not represent a significant change

in the configuration or operation of the plant. Trisodium phosphate

dodecahydrate is currently present in the containment lower level.

Design Change 3491 which increases the storage capacity of the TSP

storage baskets was evaluated in accordance with 10 CFR 50.59 and

found not to involve an unreviewed safety question.

[[Page 17235]]

Boric acid acts as a buffer to prevent the pH from rising above

approximately 8.1 as TSP is dissolved. An internal study (EC-S96-013

revision 0) has shown that given the ``ratio of grams of TSP to

liters of 3000 ppm boron solution'' stays less than 5.6, TSP cannot

increase pH above 8.2. As pH increases, components composed of

aluminum, zinc, or copper become vulnerable to corrosion. Branch

Technical Position MTEB 6-1 implies that a solution pH greater than

7.5 enhances the chance for hydrogen generation as a result of

aluminum corrosion. Waterford 3 administratively limits the amount

of aluminum in containment to minimize the amount of hydrogen

expected during a DBA. Zinc is a component of the paint applied to

surfaces inside containment. The hydrogen recombiner design basis

includes 464 square feet (1040 pounds) of aluminum and 419,300

square feet (17,252 pounds) of metallic zinc. Estimates of the

amount of hydrogen produced by the aluminum assumes that the

corrosive agent is sodium hydroxide--a much more active chemical

than is TSP. Thus, the amount of hydrogen expected in the FSAR for

the hydrogen recombiner bounds what would actually be produced by

TSP even at a pH of approximately 8.1.

The 4.5.2.d.3 proposed TSP to boron ratio assures that pH cannot

rise above 8.1 as long as post accident in-containment boric acid

solution concentration is no greater than 3011 ppm boron and no less

than 1504 ppm boron. The main variable in post accident

concentration (the difference between 1504 and 3011) is the

concentration in the RCS at the time of the accident.

Therefore, the proposed change will not create the possibility

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

evaluated.

3. Will operation of the facility in accordance with this

proposed change involve a significant reduction in a margin of

safety?

Response: No.

Trisodium phosphate dodecahydrate is stored in the containment

lower level to raise the pH of the sump and spray water following a

LOCA. As the pH of the water increases, more radioactive iodine is

kept in solution and the amount of airborne radioactive leakage is

decreased. A neutral pH also reduces the hydrogen generation from

the corrosion of the galvanized materials in containment. An

additional advantage of a higher pH is the beneficial reduction in

chloride stress corrosion cracking of metal components in the

containment following an accident.

Technical Specification 4.5.2.d.3 requires verification that a

minimum volume of TSP is contained in the storage baskets in

containment. Nine previous runs of surveillance requirement

4.5.2.d.4 (and similar tests) showed that the TSP actually used in

the plant properly neutralized a sample of water borated within RWSP

boron concentration limits. Boron concentrations of eight of the

sample solutions used in these tests ranged from 1753 ppm to 2217

ppm and resulted in a pH of 7.02 or greater. (The boron

concentration of one test performed in 1986 was unavailable.) The

ratio 4 grams to 4 liters is the amount of TSP needed to bring the

solution to a pH of at least 7.0 given that the solution is in the

1753 to 2217 ppm Boron range.

The amount of TSP in containment currently is adequate assuming

that RCS boric acid concentration stays below 454 ppm. However, the

fuel cycle is nearly over and a restart with a refreshed core would

require substantially more boric acid. We expect that the

containment water would reach approximately 2400 ppm under ideal

circumstances during cycle 9. During cycle 10, boron concentration

in containment could reach 3011 under those same ideal conditions.

As the maximum boron concentration increases, there is a non-linear

increase in the amount of TSP needed to raise solution pH to 7.0.

Thus, we request that the minimum amount of TSP in containment

required by 4.5.2.d.3 to be increased from 97.5 cubic feet to 380

cubic feet. This change also proposes to adjust the 4.5.2.d.4

specified increase that sample solution and the TSP sample size

accordingly. This change will ensure the safety injection

containment sump, when filled with water, will have an acceptable pH

following a LOCA. The test will not only demonstrate that TSP is in

the baskets but also shows that the amount of TSP in containment can

neutralize the solution expected in containment during any DBA.

Therefore, the proposed change will not involve a significant

reduction in a margin of safety. The amount of iodine kept in

solution during a DBA is limited to 50%. Note, the pH scale is

logarithmic so that the amount of TSP needed to raise pH to 7.0 is

more than three times the amount needed to reach 6.5. Furthermore,

the amount of hydrogen generated during a DBA is over estimated by

the analysis when it used sodium hydroxide as the corrosive agent

rather than TSP.

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, LA 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.

Houston Lighting & Power Company, City Public Service Board of San

Antonio, Central Power and Light Company, City of Austin, Texas, Docket

Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, Matagorda

County, Texas

Date of amendment request: May 17, 1996, as supplemented March 17,

1997.

Description of amendment request: The proposed amendment would

modify Technical Specification (TS) Section 3/4.4.5, Steam Generators,

3/4.4.6, Reactor Coolant System Leakage, and associated Bases to allow

the installation of tube sleeves as an alternative to plugging to

repair defective steam generator tubes. The proposed change would also

specify the Westinghouse topical reports to be used for sleeve design

and inspection, and identify the inspection sample size for repaired

tubes. This application was previously published in the Federal

Register on May 29, 1996, (61 FR 26938).

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). The NRC staff's review is

presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated. Listing the specific Westinghouse topical reports in the

TS binds the South Texas Project (STP) to the sleeve design and

inspection techniques identified in that revision of the topical

report. Any changes to sleeve design or inspection technique would

require a separate TS amendment.

New TS Table 4.4-3, Steam Generator Repaired Tube Inspection,

identifies the inspection sample size for steam generator tubes that

have already been repaired. This table simply identifies inspection

criteria and associated actions for repaired tubes and does not

increase the probability or consequences of an accident previously

evaluated.

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

or different kind of accident from any accident previously

evaluated. Implementation of laser welded sleeving maintains overall

tube bundle structural and leakage integrity conditions. Providing

specific Westinghouse topical report references in the TS only

serves to identify which sleeve design and inspection techniques are

being employed at STP. Likewise, the addition of Table 4.4-3

clarifies the expected inspection samples for previously repaired

tubes. The addition of Table 4.3-3 provides assurance that

previously repaired tubes will be inspected at regular intervals and

appropriate action taken if the tube is found defective. Neither of

these additions to the TS will 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. Both of these changes are being added to

clarify the STP steam generator tube inspection program and provide

more specific detail regarding steam generator tube inspection

samples and inspection techniques. By requiring inspection of

previously repaired tubes, the margin of safety is increased rather

than decreased.

[[Page 17236]]

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: Wharton County Junior College,

J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.

Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &

Bockius, 1800 M Street, N.W., Washington, DC 20036-5869.

NRC Project Director: William D. Beckner.

Houston Lighting & Power Company, City Public Service Board of San

Antonio, Central Power and Light Company, City of Austin, Texas, Docket

Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, Matagorda

County, Texas

Date of amendment request: January 28, 1997.

Description of amendment request: The proposed amendment would

relocate the details of Technical Specification (TS) Section 6.2.3 on

the Independent Safety Engineering Group (ISEG) from the Administration

Controls section of the TSs and place these details in the Updated

Final Safety Analysis Report (UFSAR) for South Texas Project, Units 1

and 2. This relocation is administrative only, and would not render any

changes to the existing plant philosophy toward the ISEG or any safety

analysis. Section 6.2.3 would be deleted from the TSs and removed from

the table of contents for Administrative Controls. Currently UFSAR

Section 13.4.2.2 describes the ISEG, but not in the detail as the

current TSs.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed changes move details from the Technical

Specifications [TSs] to the Updated Final Safety Analysis Report

(UFSAR). The changes do not result in any hardware or operating

procedure changes. The details being removed from the Technical

Specifications [TSs] are not assumed to be an initiator of any

analyzed event. The UFSAR, which will contain the removed Technical

Specification [TS] details, will be maintained using the provisions

of 10 CFR 50.59 and is subject to the change control process in the

Administrative Controls Section of the Technical Specifications

[TSs]. [In addition] any changes to the UFSAR will be evaluated per

10 CFR 50.59, no increase in the probability or consequences of an

accident previously evaluated will be allowed without prior NRC

[Nuclear Regulatory Commission] approval. Therefore, the changes do

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 changes move details from the technical

Specifications [TSs] to the Updated Final Safety Analysis Report

(UFSAR). The changes will not alter the plant configuration (no new

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

methods governing plant operation. The changes will not impose

different requirements, and adequate control of information will be

maintained. The changes will not alter assumptions made in the

safety analysis and licensing basis. Therefore, the changes will not

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

any accident previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed changes move detail from the Technical

Specifications [TSs] to the Updated Final Safety Analysis Report

(UFSAR). The changes do not reduce the margin of safety since the

relocation of details [is an administrative action and] has no

impact on any safety analysis assumptions. In addition, the detail

transposed from the Technical Specifications [TSs] to the UFSAR are

the same as the existing Technical Specification [TS] [6.2.3]. [In

addition] any future changes to the FSAR will be evaluated per the

requirements of 10 CFR 50.59, no reduction in a margin of safety

will be allowed without prior NRC approval. [Therefore, the licensee

concluded that the changes will not involve a significant reduction

in a margin of safety.]

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

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

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

request for amendments involves no significant hazards consideration.

Local Public Document Room location: Wharton County Junior College,

J. M. Hodges, Learning Center, 911 Boling Highway, Wharton, TX 77488.

Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &

Bockius, 1800 M Street, N.W., Washington, DC 20036-5869.

NRC Project Director: William D. Beckner.

Northeast Nuclear Energy Company, et al., Docket No. 50-245, Millstone

Nuclear Power Station, Unit No. 1, New London, Connecticut

Date of amendment request: February 7, 1997.

Description of amendment request: The proposed Technical

Specification changes would clarify and/or modify instrument

calibration, functional, and response time requirements for resistance

temperature detector and thermocouple testing. Also, certain

definitions would be clarified and/or modified using applicable wording

from NRC's NUREG-1433, ``Standard Technical Specifications,'' Revision

1, and industry recommendations. Additionally, the change would

relocate the reactor protection system logic response time value

utilizing the guidance provided by NRC's Generic Letter 93-08,

``Relocation of Technical Specification Tables of Instrument Response

Time Limits,'' with the exception of relocating the value to the

Technical Specifications Bases Section instead of the Updated Final

Safety Analysis Report. The proposed amendment is intended to clarify

instrumentation surveillance requirements, thereby helping to ensure

proper testing of safety-related components.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Pursuant to 10 CFR 50.92, NNECO [Northeast Nuclear Energy

Company] has reviewed the proposed changes and concludes that the

changes do not involve a significant hazards consideration (SHC)

since the proposed changes satisf[y] the criteria in 10 CFR

50.92(c). That is, the proposed changes do not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed amendment continues to ensure the surveillance

requirements satisfy the licensing basis. The current TS [technical

specifications] definition for Instrument Functional Test requires

injection of a simulated signal into the primary sensor to verify

proper response. Current TS exempt the sensors of specific

instrument channels where it is not practical to include them within

the functional test boundaries. Some examples of these exemptions

include neutron monitoring system, turbine control valve fast

closure, and standby gas treatment initiation radiation monitors. In

these cases, TS permit the performance of the functional test by

injection of a simulated electrical signal into the measurement

channel. The proposed definition, which is consistent with the STS

[standard technical specifications]

[[Page 17237]]

definition, for CHANNEL FUNCTIONAL TEST requires injection of the

simulated signal ``as close to the sensor as practicable.''

Therefore, the proposed definition is consistent with the current TS

definition and its exemptions. The primary sensor is the transmitter

or switch or radiation monitor. The definition does not include

sensing elements such as radiation detectors, flow elements,

acceleration relays or reference legs.

This change will allow the channel functional test to be

performed by means of any series of sequential, overlapping, or

total channel steps and aligns this methodology with industry

practice. This change does not affect accident precursors and thus

does not involve a significant increase in the probability of an

accident previously evaluated. The proposed change will allow a

simulated or actual signal to be used to perform an Instrument or

Channel Functional Test. This change does not impose a requirement

to create an actual signal, nor does it eliminate any restriction on

producing an actual signal. While creating an ``actual'' signal

could increase the probability of an event, existing procedures (and

the 10 CFR 50.59 control of revisions to them) dictate the

acceptability of generating this signal. The proposed change does

not affect the procedures governing plant operations or the

acceptability of creating these signals; it simply would allow such

a signal to be utilized in evaluating the acceptance criteria for

the Instrument or Channel Functional Test requirements. Therefore,

the change does not involve a significant increase in the

probability of an accident previously evaluated. Because the method

of initiation will not affect the acceptance criteria of the

Instrument or Channel Functional Test, the change does not involve a

significant increase in the consequences of an accident previously

evaluated.

Minor word differences from STS are required to provide

consistency with current TS wording and support the current

licensing basis. These minor word differences including Industry/

TSTF [Technical Specification Task Force] Standard Technical

Specification Change Traveler (TSTF-64) do not alter the meaning of

instrument testing in the STS or change the current licensing basis.

Moving the RPS [Reactor Protection System] Logic Response Time

LCO [Limiting Condition of Operation] description to the TS

definition section is an administrative change and does not alter

the original intent or licensing basis.

Relocation of the RPS Logic Response Time value from the TS to

the Bases section involves the use of an alternate regulatory

process for controlling the instrument response time limit. The

change does not introduce any new modes of plant operation, make any

physical changes, alter any operational setpoints, or change the

surveillance requirements. Any change in the RPS logic response time

value would be evaluated pursuant to the requirements of 10 CFR

50.59.

The surveillance section editorial change does not alter the

meaning of surveillance applicability. Providing RPS Logic Response

Time surveillance frequency and applicable trip functions ensures

proper testing of RPS components and is consistent with industry

practice. An evaluation completed by GE [General Electric] verified

the applicable RPS trip functions that require a specific logic

response time using the current accident analysis as the basis. For

trip functions where no explicit credit is taken in the safety

analysis, the measurement of logic response time is not important,

and therefore, not warranted. In addition, we have concluded, that

instrumentation response time requirements (specified limits) other

than RPS logic are not important to test, especially considering the

long delays already accounted for in the accident analyses

associated with the start of emergency power sources, ECCS

[Emergency Core Cooling System] components, and containment

isolations, and that the non-RPS logic response times, including

response times of other instrumentation such as radiation monitors,

are not part of the Millstone Unit No. 1 licensing basis. The

sensors associated with all TS instrumentation are functionally

tested and calibrated to ensure proper operation.

No physical change is being made to instrument channels, or to

any systems or component that interfaces with the instrumentation

channels, therefore there is no change in the probability or

consequences of any accident analyzed in the UFSAR [Updated Final

Safety Analysis Report].

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

from any accident previously evaluated.

The proposed change does not result in any design or physical

configuration changes to the instrumentation channels. Operation

incorporating the proposed change will not impair the

instrumentation channels from performing as provided in the design

basis.

Changing the TS to be consistent with current industry practice

adopted in STS will help to prevent unnecessary removal and

potential damage of the temperature detectors (for sensor

calibration). Clarification of RPS Logic Response Time testing

requirements consistent with the current licensing basis will ensure

proper testing of safety-related components.

Wording changes to Instrument Calibration and Functional Test

definitions do not involve a physical modification to the plant. The

injection of an actual or simulated signal as close to the sensor as

practical minimizes the likelihood of any transients.

Minor word differences from STS are required to provide

consistency with current TS wording and support the current

licensing basis. These minor word differences, including Industry/

TSTF Standard Technical Specification Change Traveler (TSTF-64), do

not alter the meaning of instrument testing in the STS or change the

current licensing basis.

Moving the RPS Logic Response Time LCO description to the TS

definition section is an administrative change and does not alter

the current licensing basis.

Relocation of the RPS Logic Response Time value involves the use

of an alternate process for controlling the instrument response time

limits. Therefore, the above change does not introduce any accident

initiators as it does not involve any new modes of plant operation,

make any physical changes, alter any operational setpoints, or

change the surveillance requirements.

The surveillance section editorial change does not alter the

meaning of surveillance applicability. Providing RPS Logic Response

Time surveillance frequency and applicable trip functions ensures

proper testing of RPS components and is consistent with industry

practice.

Since the proposed changes in the Technical Specifications do

not adversely impact the reliability of the RPS and other automatic

actuations, no new or different kind of accident is created.

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

Because the proposed change does not involve the addition or

modification of plant equipment, is consistent with the existing

Technical Specifications, current industry practices as outlined in

NUREG 1433, ``Standard Technical Specifications GE Plants, BWR/4,''

Revision 1, and with the current design and licensing basis of the

Protective Instrumentation systems including the accident analysis,

no action will occur that will involve a significant reduction in a

margin of safety.

The proposed change to allow the use of an actual signal in

addition to the existing requirement, which limits use to a

simulated signal, will not affect functional test acceptance

criteria. Therefore, the proposed change does not adversely affect

the reliability of the RPS or other automatic actuation and does not

involve a significant reduction in a margin of safety.

Relocation of the RPS Logic Response Time value from the TS to

the Bases section involves the use of an alternate regulatory

process for controlling the instrument response time limit. Any

change in the RPS logic response time value would be evaluated

pursuant to the requirements of 10 CFR 50.59.

Therefore, the proposed change does not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center,

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

Norwich, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49

Rope Ferry Road, Waterford, CT 06385.

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

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

CT 06141-0270.

NRC Deputy Director: Phillip F. McKee.

[[Page 17238]]

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

Nuclear Generating Plant, Wright County, Minnesota

Date of amendment request: November 25, 1996, as supplemented

December 12, 1996.

Description of amendment request: The proposed amendment would make

changes to Section 2.1.A for the Safety Limit Minimum Critical Power

Ratio (SLMCPR) and to Section 3.11.C for the Operating Limit Minimum

Critical Power Ratio (OLMCPR). The proposed change to Section 2.1.A

revises the SLMCPR value from 1.07 to 1.08 for two recirculation pump

operation and from 1.08 to 1.09 for single loop operation. The proposed

change to Section 3.11.C deletes the sentence that specifies the OLMCPR

limit penalty for single recirculation loop operation and adds a

statement that references the Core Operating Limits Report (COLR) as

the source for this information.

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 basis of the MCPR [minimum critical power ratio] Safety

Limit calculation is to ensure that greater than 99.9% of all fuel

rods in the core avoid transition boiling if the limit is not

violated. The new SLMCPRs preserve the existing margin to transition

boiling and fuel damage in the event of a postulated accident. The

probability of fuel damage is not increased. The derivation of the

revised SLMCPRs for Monticello for incorporation into the Technical

Specification, and its [their] use to determine cycle-specific

thermal limits, have been performed using NRC-approved methods as

identified in Technical Specification 6.7.A.7.b. NSP [Northern

States Power] methodology established OLMCPR such that integrity of

the SLMCPR is maintained for the bounding analyzed transients.

Additionally, GENE [General Electric Nuclear Energy] interim

implementing procedures, which incorporate cycle-specific

parameters, have been used. Based on the use of these calculations,

the calculation of the revised SLMCPRs maintains the integrity of

the safety limits and therefore cannot increase the probability or

severity of an accident. The single loop OLMCPR evaluation was

performed using NSP methodology approved by the NRC. Relocating the

OLMCPR value to the COLR establishes appropriate control on a core

operating limit which may vary from cycle to cycle because it is

cycle dependent. Since OLMCPR is developed using procedures approved

in the Technical Specifications, placing the OLMCPR in the COLR

cannot result in a change not controlled by the Technical

Specifications. The change does not affect failure modes of

equipment, therefore, this amendment will not cause a significant

increase in the probability or consequences of an accident

previously evaluated.

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

new or different kind of accident from any accident previously

analyzed.

The MCPR Safety Limit is a Technical Specification numerical

value, designed to ensure that fuel damage from transition boiling

does not occur as a result of the limiting postulated accident. It

cannot create the possibility of any new type of accident. The new

SLMCPRs have been calculated using NRC-approved methods and the

OLMCPR values are more conservative. Additionally, interim

procedures, which incorporate cycle-specific parameters, have been

used. Therefore, the proposed Technical Specification change does

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

from any accident previously evaluated.

3. The proposed amendment will not involve a significant

reduction in the margin of safety.

The MCPR Safety Limit is a Technical Specification numerical

value, designed to ensure that fuel damage from transition boiling

does not occur as a result of the limiting postulated accident.

Increasing the SLMCPR and OLMCPR values results in an increase in

the margin of safety to fuel failure, and does not affect other

plant systems. Therefore, the proposed Technical Specification

change does not involve a significant reduction in the margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401.

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

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

NRC Project Director: John N. Hannon.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun

Station, Unit No. 1, Washington County, Nebraska

Date of amendment request: November 20, 1996, as supplemented by

letter dated February 20, 1997.

Description of amendment request: The proposed amendment would

revise the technical specifications (TS) to allow the Vice President to

designate the Safety Audit and Review Committee (SARC) Chairperson, to

change the work hours limitation in accordance with guidance in GL 82-

12, ``Nuclear Power Plant Staff Working Hours;'' to change radioactive

shipments record retention requirements to comply with recent 10 CFR

Part 20 changes; to revise position titles to reflect organizational

changes; and other editorial changes. The February 20, 1997,

supplemental letter differs from the November 20, 1996, application

which was noticed in the Federal Register on January 2, 1997 (62 FR

131), in that the previous application did not propose changes to TS

5.3, 5.5, 5.6, 5.7, and 5.11 reflecting recent organizational changes.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The changes requested are administrative in nature. Paragraph

3.D was placed in the License by Amendment No. 155 to authorize

Omaha Public Power District (OPPD) to increase the storage capacity

of the FCS spent fuel pool. Amendment No. 155 stated that the TS as

issued would be effective when the last new rack was installed.

Since the last new rack was installed on August 8, 1994, Paragraph

3.D is no longer necessary and should be deleted from the License.

Table of Contents, Section 6.0, ``Interim Special Technical

Specifications,'' Subsections 6.1 through 6.4 are proposed for

deletion because all of the Specifications referred to have been

deleted by previous Amendments.

The revision proposed for TS 2.15 (Item 2C of Table 2-3 & Item

1C of Table 2-4) will insert the correct terminology (Pressurizer

Low/Low Pressure) into the Functional Unit description.

The revision proposed for TS 5.2 will delete the specific

working hours as stated and relocate these requirements to the

Updated Safety Analysis Report (USAR). Overtime will remain

controlled by plant administrative procedures with the USAR

generally following the guidance of the NRC's Policy Statement on

working hours contained in Generic Letter 82-12, ``Nuclear Power

Plant Staff Working Hours.'' Specifying personnel working hours in

TS does not meet any of the four criteria contained in 10 CFR 50.36

for inclusion in the TS. Revisions to plant procedures containing

these requirements are required to be evaluated in accordance with

10 CFR 50.59. The proposed relocation is similar to recent

Amendments issued to the Davis-Besse Nuclear Power Station and the

San Onofre Nuclear Generating Station.

The revision proposed for TS 5.5.2.2 will replace the specific

title of the Chairperson of the Safety Audit and Review Committee

[[Page 17239]]

and replace it with ``Member as appointed by the Vice President.''

This will allow the flexibility to change chairmanship of the

committee amongst the members.

The revisions proposed to TS 5.3, 5.5, 5.6, 5.7, and 5.11 revise

position titles and reporting responsibilities to reflect

organizational changes. Qualifications for individuals in these

positions meet or exceed the previous requirements.

The revision to TS 5.10 concerning retention of records of

radioactive shipments will update the TS to current 10 CFR 20

requirements. Plant procedures already comply with current 10 CFR 20

record retention requirements. The addition of the Section 5.0 title

corrects a minor format discrepancy.

These proposed revisions are administrative in nature. The

proposed revisions have no effect on any initial assumptions or

operating restrictions assumed in any accident, nor do these changes

have any effect on equipment required to mitigate the consequences

of an accident. Therefore the proposed revisions do 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 revisions correct minor errors, remove outdated

information, are consistent with changes in organizational

structure, 10 CFR Part 20, or the criteria contained in 10 CFR

50.36. These changes will not result in any physical alterations to

the plant configuration, changes to setpoint values, or changes to

the application of setpoints or limits. No new operating modes are

proposed as a result of these changes. Therefore the proposed

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

accident from any accident previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The revisions listed above correct minor errors, remove outdated

information, or are consistent with changes in organizational

structure, 10 CFR Part 20, or the criteria contained in 10 CFR

50.36. These changes will not result in any physical alterations to

the plant configuration, changes to setpoint values, or changes to

the application of setpoint or limits. 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: W. Dale Clark Library, 215

South 15th Street, Omaha, Nebraska 68102

Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L

Street, N.W., Washington, DC 20005-3502.

NRC Project Director: William H. Bateman.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo

Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,

California

Date of amendment request: February 26, 1997.

Description of amendment request: The proposed amendment would

revise the combined Technical Specifications (TS) for the Diablo Canyon

Power Plant, Unit Nos. 1 and 2 to revise TS 3/4.4.5 and 3.4.6.2,

including associated Bases 3/4.4.5 and 3/4.4.6.2, to allow the

implementation of steam generator (SG) tube voltage based repair

criteria for outside diameter stress corrosion cracking (ODSCC)

indications at tube-to-tube support plate (TSP) intersections. The

allowed primary-to-secondary operational leakage from any one SG would

be reduced from 500 gpd to 150 gpd.

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.

Structural Integrity Considerations

The structural criteria ensure that all indications subjected to

voltage-based repair limits will be able to withstand pressure

loading consistent with the criteria of NRC Regulatory Guide (RG)

1.121.

Tube burst criteria are inherently satisfied during normal

operating conditions because of the proximity of the tube support

plate (TSP). It is conservatively assumed that the entire crevice

region is uncovered during the secondary side blowdown of a main

steam line break (MSLB). Therefore, during a postulated MSLB

accident, tube burst capability must exceed the RG 1.121 criterion

requiring a margin of 1.43 times the steam line break pressure

differential on tube burst.

Based on the latest industry database, the RG 1.121 criterion is

satisfied by bobbin coil indications of outside diameter stress

corrosion cracking (ODSCC) with signal amplitudes less than 8.7

volts. The latest NRC-approved database will be used for repair and

analysis applications.

Industry testing of model boiler and operating plant tube

specimens for free-span tubing (no tube support plate (TSP)

restraint) at room temperature conditions show typical burst

pressures in excess of 5,000 psi for ODSCC indications with voltage

measurements at or below 8.7 volts. This tube burst capability

exceeds the RG 1.121 criterion.

The lower voltage repair limit is conservatively defined to be

2.0 volts in accordance with NRC Generic Letter (GL) 95-05,

``Voltage-Based Repair Criteria for Westinghouse Steam Generator

Tubes Affected by Outside Diameter Stress Corrosion Cracking,''

August 3, 1995. This 2.0 volt repair limit is very conservative

because it contains a large safety margin, based on a structural

limit of 8.7 volts. A maximum allowable upper repair limit (URL) is

also established using the guidance of GL 95-05. The URL is

calculated before each inspection by subtracting the NDE uncertainty

and growth rate allowances from the current structural limit. The

URL for near term inspections at DCPP Units 1 and 2 is expected to

be about 5.0 volts. Bobbin indications greater than 2.0 volts and

less than or equal to 5.0 volts that are confirmed by RPC will be

repaired. Bobbin indications greater than 5.0 volts will be

repaired.

Following each inspection, burst probability analyses are

performed for the end of cycle (EOC) distribution. In accordance

with GL 95-05, the projected MSLB burst probability must be less

than the threshold value of 1 x 10 x 2. Based on the relatively

small number and voltages of ODSCC indications identified to date at

DCPP Units 1 and 2, it is expected that the near term EOC

conditional burst probability for a faulted SG will be much less

than this threshold value, providing further assurance of acceptable

structural integrity.

Leakage Considerations

PG&E will implement reduced operational leakage limits as

recommended in GL 95-05. PG&E will revise the TS to implement a

maximum leakage rate of 150 gpd for any one SG to help preclude the

potential for excessive leakage during power operation in Modes 1

and 2. The TS has also been changed to specify that the 150 gpd leak

limit is not necessarily a limiting condition for operation in Modes

3 and 4. The 150 gpd leak rate per steam generator has been

established for normal operation. This leakage rate provides added

assurance against tube rupture at normal and faulted conditions. In

Modes 3 and 4, there is less differential pressure across the tube

and the potential source term from a tube failure is much less than

in Modes 1 and 2. The operational leak rate monitoring program is a

defense-in-depth measure that provides a means for identifying leaks

during power operation to allow for repair before such leaks can

result in tube failure. The leakage criteria ensure that for

indications subjected to voltage-based repair criteria, induced

leakage under worst-case MSLB conditions will not result in offsite

and control room dose releases that exceed the applicable guideline

values of 10 CFR 100 and GDC 19.

Relative to the expected leakage during accident condition

loadings, a postulated MSLB outside of containment, but upstream of

the main steam isolation valve (MSIV), represents the most limiting

radiological condition for implementation of voltage-based repair

criteria. The steam generator tubes are subjected to an increase in

differential pressure following a MSLB, resulting in a postulated

increase in leakage

[[Page 17240]]

and associated offsite doses. Leakage following a MSLB bypasses

containment.

PG&E will calculate the primary-to-secondary leakage for

degradation subjected to the voltage repair criteria under worst-

case postulated MSLB conditions. The leak rate will be compared to

the maximum allowable leak rate limit of 12.8 gpm to ensure that a

postulated MSLB occurring at EOC would not result in radiological

consequences that are in excess of the applicable offsite and

control room dose guidelines of 10 CFR 100 and GDC 19. Based on the

relatively small number of ODSCC indications identified to date at

DCPP Units 1 and 2, it is expected that the near term EOC predicted

leak rates for a faulted SG will be much less than the maximum

allowable leak rate limit.

Therefore, based on the structural integrity and leakage

considerations discussed above, the proposed changes do 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.

Implementation of the proposed voltage-based repair criteria for

ODSCC at TSP intersections does not introduce any significant change

to the plant design basis. Use of the criteria does not create a

mechanism which could result in an accident in the free span because

the repair criteria do not apply to tubes containing ODSCC located

outside the thickness of the TSPs. Based on the burst probability

acceptance limit of 1 x 10-2, it is expected that for all plant

conditions, neither a single nor multiple tube rupture event would

likely occur in a steam generator where voltage-based repair

criteria have been applied.

Steam generator tube integrity is continually maintained through

inservice inspection and primary-to-secondary leakage monitoring.

Any tubes with ODSCC degradation in excess of the URL are repaired.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The use of the bobbin probe to disposition ODSCC degraded tubes

within TSP intersections by voltage-based repair criteria is

demonstrated to maintain SG tube integrity in accordance with the

requirements of RG 1.121. RG 1.121 describes a method acceptable to

the NRC Staff for meeting GDCs 14, 15, 31, and 32 by reducing the

probability or the consequences of SG tube rupture. This is

accomplished by determining the limiting conditions of degradation

of SG tubing, as established by inservice inspection, for which

tubes with unacceptable degradation are removed from service. Upon

implementation of the voltage-based repair criteria, even under the

worst case conditions, the occurrence of ODSCC at TSP intersections

is not expected to lead to a SG tube rupture during normal or

faulted plant conditions, nor is it expected to lead to unacceptable

primary-to-secondary leakage.

In addressing the combined effects of a loss of coolant accident

(LOCA) and safe shutdown earthquake (SSE) on the SGs, as required by

GDC 2, it has been determined that tube collapse may occur based on

analysis for a large break LOCA plus SSE. The analysis identifies a

maximum of 7.5 percent of tubes per SG located adjacent to wedge

regions that are subject to potential collapse during combined LOCA

and SSE. Tubes located in the wedge region exclusion zone will be

excluded from application of voltage-based repair criteria. Thus,

existing tube integrity requirements apply to these tubes and the

margin of safety is not reduced.

Implementation practices using voltage-based repair criteria

bounds RG 1.83 considerations. Specifically, GL 95-05 requires the

following: (1) enhanced eddy current inspection guidelines are

implemented to provide consistency in voltage normalization; (2) 100

percent bobbin coil inspections are performed each cycle for all hot

leg TSP intersections and all cold leg TSP intersections down to the

lowest cold leg TSP with known ODSCC indications; and (3) rotating

pancake coil (RPC) inspection of indications greater than 2 volts

are performed to characterize the principal degradation as ODSCC.

DCPP's proposed voltage-based repair criteria implementation

practices meet the above requirements, and in some areas exceed them

(for example, 100 percent bobbin coil inspections are routinely

performed each cycle on every TSP intersection).

Implementation of voltage-based repair criteria at TSP

intersections will decrease the number of tubes which must be

repaired. Since the installation of tube plugs to remove ODSCC

degraded tubes from service reduces RCS flow margin, voltage-based

repair criteria implementation will help preserve the margin of RCS

flow.

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 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: California Polytechnic State

University, Robert E. Kennedy Library, Government Documents and Maps

Department, San Luis Obispo, California 93407.

Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and

Electric Company, P.O. Box 7442, San Francisco, California 94120.

NRC Project Director: William H. Bateman.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323,

Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis

Obispo County, California

Date of amendment requests: February 27, 1997.

Description of amendment requests: The proposed amendments would

revise the combined Technical Specifications (TS) for the Diablo Canyon

Power Plant, Unit Nos. 1 and 2 by revising Technical Specifications

(TS) 3/4.8.1.1, ``A.C. Sources--Operating,'' to clarify that emergency

diesel generator (EDG) testing is initiated from standby conditions

rather than ``ambient'' conditions. The associated TS Bases will be

revised to discuss the temperature range that satisfies EDG standby

conditions. This amendment also proposes to revise TS 3/4.3.2,

``Instrumentation--Engineering Safety Features Actuation System

Instrumentation.'' This revision clarifies that when one or both of the

first level load shed relays, or one or both of the second level

undervoltage relays are inoperable, the associated EDG for that bus

shall be declared inoperable.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed changes to the technical specifications (TS) do not

change the function or operation of any plant equipment or affect

the response of that equipment if it is called upon to operate.

The proposed change to TS 4.8.1.1.2a.2 and the Bases will

clarify the term ``ambient conditions'' as used in the emergency

diesel generator (EDG) surveillance requirement. EDG testing will

still be completed on a frequency commensurate with the current TS.

The proposed change to TS 3.3.2, Table 3.3-3, will permit time

to restore the load shed first level undervoltage relays (FLURs) and

second level undervoltage relays (SLURs) to operable status that is

consistent with times allowed for outage of other safety-related

equipment affecting one train of vital equipment. This proposed

change maintains a high degree of equipment availability without

requiring unnecessary initiation of a plant shutdown for partial

equipment outages.

Therefore, the proposed changes do 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

[[Page 17241]]

accident from any accident previously evaluated.

The proposed change to TS 4.8.1.1.2a.2 and the Bases will

clarify the term ``ambient conditions'' as used in the EDG

surveillance requirement. EDG testing will still be completed on a

frequency commensurate with the current TS, and will be more

representative of the conditions under which the EDGs would be

required to start in an accident condition.

The proposed change to TS 3.3.2, Table 3.3-3, will provide time

to restore the load shed FLURs and SLURs to operable status that is

consistent with times allowed for outage of other safety-related

equipment affecting one train of vital equipment. The load shed FLUR

and SLUR sets for one 4 kV bus only affect one train of vital

equipment. If an accident occurred while the relays were inoperable,

the redundant trains (two remaining EDGs and vital buses) would

complete the safety function. The proposed allowed outage time (AOT)

for the load shed FLURs and SLURs is bounded by the time allowed for

an EDG supporting the vital 4 kV bus and is consistent with AOTs for

other safety-related components.

Therefore, the proposed changes do not create the possibility of

a new or different kind of accident from any accident previously

evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed change to TS 4.8.1.1.2a.2 and its Bases, clarifies

the term ``ambient conditions'' as used in the EDG surveillance

requirement. EDG testing will still be completed on a frequency

commensurate with the current TS. Use of temperatures in the standby

range result in no significant variation in EDG start times as

indicated by the diesel vendor and by PG&E test results. Standby

conditions are representative of actual starting conditions that

would be in effect if the EDGs started in an accident.

The proposed change to TS 3.3.2, Table 3.3-3, will provide time

to restore the load shed FLURs and SLURs to operable status that is

consistent with times allowed for outage of other safety-related

equipment affecting one train of vital equipment. If an accident

occurred while the relays were inoperable, the redundant trains (two

remaining EDGs and vital buses) would complete the safety function.

The proposed change eliminates an unneccessary plant shutdown and

associated risk due to shutdown transient. It prevents a transient

that could require the EDGs at a time when less than all three EDGs

would be operable.

Therefore, neither of the proposed changes involves a

significant reduction in a margin of safety.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: California Polytechnic State

University, Robert E. Kennedy Library, Government Documents and Maps

Department, San Luis Obispo, California 93407.

Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and

Electric Company, P.O. Box 7442, San Francisco, California 94120.

NRC Project Director: William H. Bateman.

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

Trojan Nuclear Plant, Columbia County, Oregon

Date of amendment request: January 28, 1997.

Description of amendment request: The proposed amendment by

Portland General Electric (PGE or the licensee) clarifies the

administrative controls that are used for the revision and maintenance

of the Certified Fuel Handler Training Program. The change allows the

licensee to make changes to the certified fuel handlers program without

prior NRC staff approval. The text of the proposed change is taken from

the improved standard technical specifications, NUREG-1431, ``Standard

Technical Specifications, Westinghouse Plants.''

Basis for proposed no significant hazards consideration

determination: As required by 10 CFR 50.91(a), the licensees have

provided their analysis of the issue of no significant hazards

consideration, which is presented below:

In accordance with the requirements of 10 CFR 50.92, ``Issuance

of amendment,'' this license amendment request is judged to involve

no significant hazards consideration based upon the following:

1. The proposed license amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed change is a clarification of the method of control

that will be used for the Certified Fuel Handler Training Program,

and as such, is administrative in nature and has no impact on the

probability or consequences of accidents previously evaluated. The

physical structures, systems, and components of the facility and the

operating procedures for their use are unaffected by this proposed

clarification. The proposed administrative controls provide adequate

confidence that personnel that perform the certified fuel handler

functions will have been adequately trained for the changing

conditions of the facility. Since the training program will prepare

the operations personnel for fuel handling operations, including

responses to abnormal events/accidents, there will be no increase in

the probability of occurrence or in the consequences of an accident

previously evaluated.

2. The proposed license amendment does not create the

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

previously evaluated.

This change ensures the qualifications of the operations

personnel are commensurate with the tasks to be performed and the

conditions to which they may be required to respond. This change

does not affect plant equipment or the procedures for operating

plant equipment and, therefore, does not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

3. The proposed license amendment does not involve a significant

reduction in a margin of safety.

This change ensures the qualification of the operations

personnel are commensurate with the tasks to be performed and the

conditions to which they may be required to respond. The assumptions

for a fuel handling accident in the Fuel Building are not affected

by the proposed change. The proposed amendment does not, therefore,

involve a reduction in a margin of safety.

The NRC staff has reviewed the analysis of the licensee 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: Branford Price Millar Library,

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

Portland, Oregon 97207.

Attorney for the Licensees: Leonard A. Girard, Esq., Portland

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

97204.

NRR Project Director: Seymour H. Weiss.

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271,

Vermont Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: September 11, 1996.

Description of amendment request: The proposed amendment would

permit operation with increased safety relief valve (SRV) and safety

valve (SV) setpoint tolerance and permit operation up to 100% of rated

power with a single inoperable SRV.

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 changes will permit operation with increased SRV

and SV setpoint tolerance and permit operation up to 100% of rated

power with a single inoperable SRV.

[[Page 17242]]

The valves are not related to the control rod system. The valves

are not involved in the initiation of a Control Rod Drop Accident.

The valves are part of the Reactor Vessel (RV) pressure boundary and

their failure could initiate a LOCA [loss-of-coolant accident].

However, the proposed changes do not constitute a change in the

design of the valves from a pressure boundary perspective. The

proposed changes do not affect the probability of a LOCA initiated

by valve failure. The valves are not a component, system, or

structure involved in refueling operations. The valves and their as-

found setpoint tolerance are not involved in the initiation of a

Refueling Accident.

The design basis Main Steam Line Break is a complete severance

of one main steam line outside the secondary containment. The SRVs

and SVs are located inside primary containment and cannot cause a

main steam line rupture outside secondary containment. The valves

are not involved in the initiation of a design basis Main Steam Line

Break. The probability or consequences of these accidents are not

affected.

Attachment C [see application dated September 11, 1996] includes

an analysis to demonstrate that margin exists to SV challenges

during an Abnormal Operational Transient (AOT). For this purpose a

Generator Load Rejection without Bypass (GLRWOBP) was identified as

the limiting AOT. The results confirm that SV challenges would not

occur with an inoperable SRV at rated power.

The current Technical Specification limit of 95% rated power or

less with an inoperable SRV is therefore not required to prevent SV

challenges during an AOT.

As discussed in Attachment C [see application], the impact of

the proposed as-found SRV setpoint tolerance increase on SRV piping/

supports and discharge loads to the Torus was evaluated. A

mechanical loads analysis confirmed the integrity of these

components, systems, and structures during SRV discharge with the

proposed changes.

Attachment C [see application] provides an evaluation of the

impact of the proposed changes on the consequences of the Loss of

Coolant Accident and the Main Steam Line Break. The limiting LOCA

event is a break in the recirculation loop, with a break area of 0.6

ft\2\, at the pump discharge location, with a loss of one train of

DC power as the single failure. For breaks in the recirculation line

larger than 0.4 ft\2\, the SRVs would not be challenged. Therefore,

in assessing the impact of the proposed changes on 10CFR50.46

acceptance criteria, only recirculation line breaks less than 0.4

ft\2\ were reevaluated. Results show that the 0.6 ft\2\

recirculation line break remains the limiting LOCA event and it is

not affected. The consequences of the limiting design basis LOCA are

not increased by the proposed changes. The design basis accident for

containment performance is a double-ended break in the recirculation

pump suction. For this size break, the SRVs are not challenged.

Therefore, the proposed changes do not have any effect on the design

basis accident for containment performance. The design basis

accident for radioactive material releases and radiological effects

is a complete severance of one main steam line outside the secondary

containment. For steam line breaks outside the containment, MSIVs

[main steam isolation valves] close and terminate radiological

releases outside the containment, SRVs are not challenged until

after MSIV closure and isolation. Therefore, the proposed changes do

not increase the radiological consequences of the design basis Main

Steam Line Break.

The SRVs and SVs are designed to mitigate the consequences of

malfunctions of equipment which result in a Nuclear System pressure

increase. These abnormal operational transients are defined and

analyzed in Section 14.5.1 of the VY [Vermont Yankee] FSAR [final

safety analysis report]. The impact

This text is long and has been trimmed here. Open the source document for the complete record.