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. Pub. L. 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 April 10 through April 24, 1998. The last

biweekly notice was published on April 22, 1998 (63 FR 19964).

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.

[[Page 25102]]

However, should circumstances change during the notice period such that

failure to act in a timely way would result, for example, in derating

or shutdown of the facility, the Commission may issue the license

amendment before the expiration of the 30-day notice period, provided

that its final determination is that the amendment involves no

significant hazards consideration. The final determination will

consider all public and State comments received before action is taken.

Should the Commission take this action, it will publish in the Federal

Register a notice of issuance and provide for opportunity for a hearing

after issuance. The Commission expects that the need to take this

action will occur very infrequently.

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

Directives Branch, Division of Administration Services, Office of

Administration, U.S. Nuclear Regulatory Commission, Washington, DC

20555-0001, and should cite the publication date and page number of

this Federal Register notice. Written comments may also be delivered to

Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,

Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of

written comments received may be examined at the NRC Public Document

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

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

is discussed below.

By June 5, 1998, the licensee may file a request for a hearing with

respect to issuance of the amendment to the subject facility operating

license and any person whose interest may be affected by this

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

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

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

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

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

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

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

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

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

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

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

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

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

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

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

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

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

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

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

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

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

[[Page 25103]]

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 amendment request: November 1, 1996, as supplemented by

letters dated October 13, 1997, February 26, 1998, and March 13, 1998.

Description of amendment request: Associated with a Carolina Power

& Light Company (the licensee) application to convert from the Current

Technical Specifications (CTS) for the Brunswick Steam Electric Plant,

Units 1 and 2, to Improved Technical Specifications (ITS), as contained

in Revision 1 of NUREG-1433, ``Standard Technical Specification General

Electric Plants, BWR/4,'' the licensee proposed removing a restriction

on a surveillance test described below.

CTS 4.8.1.1.1.b requires that the offsite electrical power circuits

be demonstrated OPERABLE, at least once per 18 months during shut down,

by manually transferring the unit power supply from the normal circuit

to the alternate circuit. As proposed, ITS SR 3.8.1.8.b will not

contain the restriction to perform the Surveillance ``during

shutdown.'' Currently, this test is performed by momentarily

paralleling the 230 kV offsite alternating current (AC) power sources.

The licensee has stated that paralleling offsite AC power sources is a

controlled evolution and the increased risk associated with the

performance of this test while the unit is at power is not significant

for the following reasons: (1) the frequency and voltages are verified

to be within the required range prior to paralleling the two offsite AC

power sources; (2) breaker interlocks ensure that the alternate circuit

is connected to the load prior to opening the preferred circuit; (3)

the test does not result in de-energization of any 4.16 kV emergency

bus and the potential for electrical perturbations on the grid system

is the same whether performing the transfer while the unit is at power

or while shutdown; and (4) operating history indicates that

transferring offsite AC power sources while the units were in

Operational Conditions 1 (power operation) or 2 (startup) has been

performed satisfactorily without electrical distribution system

perturbations. The licensee has further pointed out that Generic Letter

91-04, ``Changes in Technical Specifications to Accommodate a 24-Month

Fuel Cycle,'' states that licensees may omit the Technical

Specification qualification that a refueling interval surveillance is

to be performed ``during shutdown.'' Therefore, consistent with the

guidance provided in Generic Letter 91-04, the licensee proposed

deletion of the requirement to perform this Surveillance ``during

shutdown'' as part of the conversion from CTS 4.8.1.1.1.b to ITS SR

3.8.1.8.b.

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?

This change would remove a specific restriction to perform the

verification of the manual transfer of the unit power supply from

the normal circuit to the alternate circuit ``during shutdown.'' The

transfer of the unit power supply from the normal circuit to the

alternate circuit is not an initiator of any previously analyzed

accident. Therefore, this change does not significantly increase the

frequency of such accidents. Currently, this test is performed by

momentarily paralleling the 230 kV offsite AC power sources.

Paralleling offsite AC power sources is a controlled evolution and

the increased risk associated with the performance of this test

while the unit is at power is not significant for the following

reasons: (1) The frequency and voltages are verified to be within

the required range prior to paralleling the two offsite AC power

sources; (2) breaker interlocks ensure that the alternate circuit is

connected to the load prior to opening the preferred circuit; (3)

the test does not result in de-energization of any 4.16 kV emergency

bus and the potential for electrical perturbations on the grid

system is the same whether performing the transfer while the unit is

at power or while shutdown; and (4) operating history indicates that

transferring offsite AC power sources while the units were in MODE

(Operational Condition) 1 or 2 has been performed satisfactorily

without electrical distribution system perturbations. The

appropriate plant conditions for performance of the Surveillance

will continue to be controlled to assure the potential consequences

are not significantly increased. This control method has been

previously determined to be acceptable as indicated in Generic

Letter 91-04. Therefore, this change does not significantly increase

the consequences of any previously analyzed accident.

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

kind of accident from any accident previously evaluated?

This change removes a specific restriction on the plant

conditions for performing a Surveillance, but does not change the

method of performance. The appropriate plant conditions for

performance of the Surveillance will continue to be controlled to

assure the possibility for a new or different kind of accident are

not created. This control method has been previously determined to

be acceptable as indicated in Generic Letter 91-04. Therefore, this

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

accident from any previously analyzed accident.

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

of safety?

The margin of safety considered in determining the appropriate

plant conditions for performing the Surveillance will continue to be

controlled to assure that there is no significant reduction. This

control method has been previously determined to be acceptable as

indicated in Generic Letter 91-04. Therefore, the 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: 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: Pao-Tsin Kuo

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 amendment request: April 3, 1998.

Description of amendment request: The Carolina Power & Light

Company, licensee for the Brunswick Steam Electric Plant (BSEP), Unit

Nos. 1 and 2, proposed amendments to the Technical Specifications (TS)

to change the specified total volume of the condensate storage tank

(CST) from 150,000 gallons to 228,200 gallons. During a recent review

of industry operating experience, the licensee determined that

information contained in TS 3.5.3.1, Core Spray System (CSS), and the

associated bases regarding water inventory in the CST was incorrect.

Specifically, the minimum CST volume requirement contained in TS

3.5.3.1 would not assure the availability of 50,000 gallons of water

for the CSS, as indicated in TS Bases section 3/4.5.3.1 for the CSS.

The licensee has concluded that the proposed license amendments do

not involve a Significant Hazards Consideration. In support of this

determination, an evaluation of each of the three standards set forth

in 10 CFR 50.92 is provided below.

[[Page 25104]]

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 license amendments do not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed TS change revises the minimum CST [Condensate

Storage Tank] water volume required for OPERABILITY of the Core

Spray system (CSS) in OPERATIONAL CONDITIONS 4 AND 5 when the

suppression pool is inoperable. The proposed change does not alter

the operation of any plant system or component; does not involve a

physical modification to any structure, system, or component; and

does not affect an initiator to any accident previously evaluated.

The minimum CST water level is being increased to assure the

availability of 50,000 gallons of water for use by the CSS.

Therefore, the proposed license amendments do not involve an

increase in the probability or consequences of an accident

previously evaluated.

2. The proposed license amendments will not create the

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

previously evaluated. This proposed TS change revises the minimum

CST water volume required for OPERABILITY of the CSS in OPERATIONAL

CONDITIONS 4 and 5 when the suppression pool is inoperable. The

proposed change does not alter the operation of any plant system or

component; does not involve a physical modification to any

structure, system, or component; and does not affect an initiator to

any accident previously evaluated. The proposed change does not add

or modify equipment or components related to the CSS and will,

therefore, not create new failure modes or common failure modes. The

minimum CST water level is being increased to assure the

availability of 50,000 gallons of water for use by the CSS.

Therefore, the proposed license amendments 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 proposed license amendments

increase the minimum CST water level to assure the availability of

50,000 gallons of water for use by the CSS. These volumes ensure the

validity of existing analyses, and ensure that the existing TS Bases

are satisfied. The proposed change does not involve a physical

modification to any structure, system, or component, and does not

modify the operation of any existing equipment. Therefore, the

proposed license amendments do not involve a reduction in a margin

of safety.

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

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

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

amendments 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: Pao-Tsin Kuo (Acting)

Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad Cities

Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois

Date of amendment request: March 31, 1998.

Description of amendment request: Unreviewed Safety Question

involving use of Station Blackout (SBO) diesel generators (DGs) and use

of a mobile safe shutdown (SSD) battery cart in the 10 CFR part 50,

appendix R, Safe Shutdown Safety 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:

The licensee has provided a separate no significant hazards

consideration determination for the SBO DGs and the battery cart under

this amendment request. The following is the determination for the SBO

DGs:

(1) No significant increase in the probability or consequences

of an accident previously evaluated is involved because of the

following:

Two types of previously evaluated accidents are relevant to this

criterion: (1) A fire; (2) other accident evaluated in the UFSAR.

For these previously evaluated accidents, the change would not

result in an increase in either their probabilities of occurrence or

the consequences of their occurrence, for the following reasons.

The use of the SBO DGs in lieu of the [Emergency Diesel

Generators] EDGs does not change the probability or consequences of

a fire. The likelihood of a fire is unchanged. Use of the SBO DGs

does not significantly change the fire loading nor introduce

significant new ignition sources. The consequences of a fire are

unchanged because use of the SBO DGs continues to support the

station's ability to achieve and maintain shutdown in the event of a

fire.

Use of the SBO DGs for non-fire purposes is unchanged by use of

the SBO DGs for post-fire safe shutdown in the event of a fire in

areas requiring alternate shutdown capability. Accordingly there is

no change in the probability or consequences of a previously

evaluated accident involving the SBO DGs. Similarly, there is no

change to the probability or consequences of other accidents that

have been previously evaluated because they are independent of this

change in use of the SBO DGs.

(2) The possibility of a new or different kind of accident from

any accident previously evaluated is not created because:

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

different kind of accident from that previously evaluated for Quad

Station. Although the SBO DGs will be used for a new function, there

is no significant change in the operation of the SBOs for a non-fire

event. Moreover, the overall use of the SBO DGs as an AC power

source is not significantly different from the use of the EDGs. The

SBO DGs buses provide power to the same buses that are powered from

the EDGs. No new modes of operation are introduced by the proposed

changes. The use of the SBO DGs provides a slightly different but

effective method for achieving and maintaining post-fire safe

shutdown for areas requiring alternate shutdown capability. As such,

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

different kind of accident.

(3) No significant reduction in the margin of safety is involved

because:

A change in the fire protection program does not result in a

significant reduction in the margin of safety if the change does not

result in a significant adverse impact on the plant's ability to

achieve and maintain safe shutdown in the event of a fire. The

proposed use of the SBO DGs instead of the EDGs to achieve and

maintain safe shutdown within 72 hours change does not significantly

affect the capability or reliability of the equipment assumed to

operate in the safety analysis.

The demonstrated capability and reliability of the SBO and EDGs

are not significantly different. Indeed, the SBO DGs represent a

safety improvement due to their physical separation from the

postulated fire areas, and the operational benefits provided by

their greater capacity. Any narrow reduction in margin associated

with the need to manually start the SBO DGs is offset by the

reduction in manual actions necessary to reduce electrical loads

powered from the EDGs. The lack of Class 1E qualification for the

SBO DGs is not significant from a safety perspective because the

demonstrated reliability of the SBO DGs is comparable to the

reliability of the EDGs. The lack of seismic qualification and

single failure protection do not constitute a significant reduction

in margin since neither of these attributes is required by Appendix

R. Accordingly, the Commission has already determined that these

attributes are not part of the Appendix R acceptance criterion. Any

reduction in margin associated with the greater fuel consumption

rate of the SBO DGs is partially offset by the increased flexibility

in powering equipment to achieve and maintain post fire safe

shutdown. Additionally, onsite fuel storage and manual transfer

capabilities provide for at least 72 hours of SBO DG operation.

Within 72 hours, deliveries of diesel fuel from offsite supplies is

expected. Therefore, the use of the SBO DGs as an onsite AC power

source for

[[Page 25105]]

equipment necessary to achieve and maintain post-fire safe shutdown

in areas requiring alternate capabilities does not involve a

significant reduction in margin.

The licensee has evaluated the use of the mobile SSD battery cart to

provide the power source for the Automatic Depressurization System

(ADS) valves under certain scenarios where the valves are needed to

achieve cold shutdown and determined that it does not involve a

significant hazards consideration for the reasons discussed below.

(1) No significant increase in the probability or consequences

of an accident previously evaluated is involved.

The accident previously evaluated is the postulated fire

requiring alternate shutdown capability. The probability of a

previously evaluated fire is not increased significantly because the

mobile SSD batteries do not create significant new ignition sources

or any other fire initiators. The consequences of a previously

evaluated fire are not increased significantly because the mobile

SSD batteries do not significantly increase the fire loading in the

plant, do not interfere with the plant's ability to extinguish a

fire, and are fully capable of fulfilling the designed safety

function.

The associated systems related to this proposed change are not

affected in a way that could impact the initiation of any accident

sequence for the Quad Cities Station. No modes of operation are

introduced by the proposed change such that adverse consequences

result.

The probability of an accident involving the use of the mobile

SSD batteries would not be increased significantly by this proposed

use because the use is not significantly different from the

alternative manual attachment of a power source to the ADS valves.

The consequences of an accident involving the use of the mobile

SSD batteries are not increased because the only significant

consequences would be a delay in achieving cold shutdown and that

would have no different consequences than would a delay due to an

accident related to the currently used manual power source.

(2) The possibility of a new or different kind of accident from

any accident previously evaluated is not created.

The proposed change for the Quad Cities Station does not create

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

previously evaluated. Because the mobile SSD batteries simply

provide a different form of manually connecting a source of power to

the ADS valves, the use of the mobile SSD batteries does not present

new or different kinds of accidents related to such manual actions.

Finally, because no new modes of operation are introduced by the

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

or different kind of accident that could be related to new modes of

operation.

(3) No significant reduction in the margin of safety is

involved.

The analytic framework for determining the extent to which a

proposed change affects the margin of safety has been discussed

above and, so will not be repeated here. In this case, a review of

the proposed changes shows that they will not have an adverse impact

on the ability to achieve and maintain safe shutdown. Several

features associated with the use of the mobile SSD batteries show,

as discussed above, that it provides an effective method for

achieving and maintaining safe shutdown following a fire. In

particular, use of the mobile SSD batteries reduces the overall

complexity of the cold shutdown repairs required to supply power to

the ADS valves and is familiar to plant personnel from their

training on its use for other purposes.

Design calculations regarding capabilities of the mobile SSD

batteries show they will be capable in fulfilling their intended

safety function for their design basis Appendix R scenario.

Reliability of the mobile SSD batteries will be maintained by

augmented quality standards. This will entail the conduct of

appropriate maintenance and surveillance which is designed to ensure

that the mobile batteries will function as intended. Reliability of

this power source is further enhanced by the circumstance that there

are two mobile SSD batteries, thus permitting one to act as a backup

to the other.

Under these circumstances, the margin of safety for achieving

cold shutdown using the ADS valves is not reduced significantly, if

at all, by the use of non-safety related mobile SSD batteries to

power the ADS valves. Although safety-related station batteries had

previously been used in this function, the method for attaching

those batteries was more prone to human error than the method which

has been developed for the mobile SSD batteries. Moreover,

substantial steps have been taken to provide a high level of

reliability for the mobile SSD batteries. Overall, therefore, the

ability to achieve and maintain safe shutdown in the event of a fire

has not been reduced by this change in the source of power to the

ADS valves.

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 are

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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Dixon Public Library, 221 Hennepin

Avenue, Dixon, Illinois 61021

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

One First National Plaza, Chicago, Illinois 60603

NRC Project Director: Stuart A. Richards

Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion

Nuclear Power Station, Units 1 and 2, Lake County, Illinois

Date of amendment request: March 30, 1998.

Description of amendment request: The proposed amendments would

restore the Zion Custom Technical Specifications (CTS) that had been

replaced with Improved Technical Specification by a previous amendment

and would reinstate License Conditions that were deleted by that

previous amendment. The proposed amendment would also modify the CTS to

allow the use of Certified Fuel Handlers to satisfy shift staffing

requirements and would change management titles and responsibilities to

reflect the permanently shutdown organization.

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:

Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

With a plant permanently shutdown and defueled the spectrum of

accidents and events that remain credible is significantly reduced.

As discussed below the proposed changes do not affect the

probability or consequences of any accidents that do remain

credible.

The restoration of the CTS which were replaced with the ITS by

Amendments 178/165 cannot increase the probability or consequences

of any event or accident because the amendment was never

implemented. The CTS have been maintained as the legally binding

Technical Specifications in effect at Zion Station. The

reinstatement of the five License Conditions deleted by Amendments

178/165 is an administrative change in that the requirements

contained in the License Conditions had been relocated elsewhere and

are now being restored exactly as they were before the amendment was

issued. Since the actual requirements have not changed there can be

no change in the probability or consequences of any accident or

event.

The changes in management titles and responsibilities will not

increase the probability or consequences of any accident or event

because these changes are administrative and will not result in any

decrease in the quality of management applied to Zion Station. The

changes are commensurate with the significant reduction in site

activities, site staffing, and risk to public health and safety that

occurs when an operational nuclear power plant transitions to a

permanently shutdown and defueled plant. Responsible individuals

will have the authority to commit the personnel and resources

necessary to fulfill their obligations for safe storage and handling

of nuclear fuel. The change of position designations will have no

effect on the frequency of occurrence of accident or event

initiators, or on their consequences.

The changes to allow use of Certified Fuel Handlers in lieu of

personnel licensed in accordance with 10 CFR part 55 will not

increase the probability or consequences of an accident or event

because the Certified Fuel Handler Training and Retraining program

(which will be approved by the

[[Page 25106]]

NRC) has been developed using a Systems Approach to Training as

defined in 10 CFR 55.4. This approach provides assurance that the

Certified Fuel Handlers have the knowledge, skills, and abilities

that are commensurate with the tasks to be performed (i.e., the

proper monitoring, handling, storage, and cooling of nuclear fuel).

Therefore the frequency of occurrence of accident or event

initiators is not increased and the consequences of the accidents or

events are unaffected.

The changes in shift staffing numbers and crew composition will

not increase the probability or consequences of an accident or

event. These staffing changes are commensurate with the quantity,

complexity, and hazard level of the activities required for storage

and handling of nuclear fuel. The elimination of the Shift Control

Room Engineer does not affect any accident or event initiator or

consequence since the previous specification would not have required

that the position be manned with both units shut down. The

elimination of the requirement for a Radiation Protection Person on

shift will have no effect on the frequency of occurrence of

accidents or events, nor on the consequences of the accident or

event.

The changes in verbiage to eliminate any implication that units

are operational will not increase the probability or consequences of

an accident or event because they are largely editorial changes and

do not increase the frequency of occurrence of [or] event

initiators, nor do they increase the consequences.

Therefore this proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

Does the change create the possibility of a new or different

kind of accident from any accident previously evaluated?

The changes proposed by this amendment do not involve new

structures, systems, or components, or the use of existing

structures, systems, or components in a new manner. Consequently no

new failure mechanisms are introduced. The design and operation of

structures, systems, or components is unaffected by:

The restoration of CTS,

The reinstatement of the five License Conditions deleted by

Amendments 178/165,

The changes in management titles and responsibilities,

The changes to allow use of Certified Fuel Handlers in lieu of

10 CFR [Part] 55 licensed personnel,

The changes in shift staffing numbers and crew composition, or

The changes in verbiage to eliminate any implication that units

are operational.

Therefore, the proposed changes do not create the possibility of

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

Does the change involve a significant reduction in a margin of

safety?

One of the License Conditions that would be reinstated by this

amendment establishes limits that help ensure that the assumptions

of the fuel handling accident analysis remain valid. License

Condition 2.C.(7).b limits the weight of loads carried over fuel

stored in the spent fuel pool to the weight of a single fuel

assembly plus the tool for moving that assembly. This weight limit

ensures that the number of fuel rods broken in a fuel handling

accident does not exceed the maximum number of fuel rods assumed to

break in the accident analysis. Consequently, this change continues

to provide assurance that the margin of safety involving the number

of fuel rods broken in the accident will not be reduced.

Therefore, these changes do not involve a significant reduction

in the margin of safety.

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

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

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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Waukegan Public Library, 128 N.

County Street, Waukegan, Illinois 60085

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

One First National Plaza, Chicago, Illinois 60603

NRC Project Director: Stuart A. Richards

Duke Energy Corporation (DEC), et al., Docket Nos. 50-413 and 50-414,

Catawba Nuclear Station, Units 1 and 2, York County, South Carolina

Date of amendment request: May 27, 1997, as supplemented by a

letter dated April 20, 1998.

Description of amendment request: The proposed amendments would

revise the Technical Specifications (TS) of each unit to conform with

NUREG-1431, Revision 1, ``Standard Technical Specifications--

Westinghouse Plants.'' The Commission had previously issued a Notice of

Consideration of Issuance of Amendments in the Federal Register on July

14, 1997 (62 FR 37628) covering all the proposed changes that were

indeed within the scope of NUREG-1431. In DEC's May 27, 1997,

submittal, there are proposed changes that are beyond the scope of

NUREG-1431, which were thus not covered by the staff's July 14, 1997,

notice. The following descriptions and no significant hazard analyses

cover only those beyond-scope changes. Associated with each change are

administrative/editorial changes such that the new or revised

requirements would fit into the format of NUREG-1431.

1. This proposed change affects the surveillance requirement

currently contained in Sections 4.6.6.1 and 4.6.6.2, regarding the

containment valve injection water system. The requirement to assure

adequate capacity to maintain system pressure for at least 30 days

would be deleted, the required system pressure of 16.2 pounds per

square inch gauge (psig) would be replaced with a surge tank pressure

of 36.4 psig, and the system would be tested at lower pressures and

more restrictive leak rates.

2. Section 3.9.2.1, regarding the boron dilution mitigating system,

currently requires both trains to be operable in Mode 6 (refueling).

DEC proposed to add a note stating that the system may be blocked

during core reloading until two assemblies are loaded into the core.

Adequate shutdown margin will continue to be controlled and verified by

other specifications. This blocking would prevent inadvertent actuation

of the system, which could distract the operating personnel, but would

not diminish the monitoring function of the system.

3. DEC proposed to change the definition of `dose equivalent

iodine-131.' Subsequently, this proposed change was withdrawn by letter

dated April 20, 1998.

4. DEC proposed to change Section 3.3.3.6 regarding accident

monitoring instrumentation. Specifically, the change would (a) increase

the time allowed to return the required number of channels to operable;

and (b) permit continued operation if one channel is inoperable given

certain conditions are met, instead of requiring shutdown.

5. DEC proposed to change Section 4.6.4.1 regarding surveillance

requirements for the hydrogen monitors (combustible gas control).

Specifically, this would eliminate the channel operational test, and

extend the channel check frequency from once per 12 hours to once per

31 days.

6. DEC proposed to change Section 3.4.6.1 regarding reactor coolant

leakage detection systems; a system comprising diverse instruments such

as gaseous radioactivity monitoring, containment floor and equipment

sump monitoring, etc. In addition to the instruments specified by this

section, the plant has other installed instruments such as monitors for

humidity, temperature, etc., which can provide indication for reactor

coolant leakage. Currently, this specification allows operation up to

30 days if the containment floor and equipment sump monitoring system

is inoperable. The change would impose a requirement to perform a

precision water balance of the reactor coolant system every 24 hours

during this period. The change would also reduce the number of monitors

required operable provided compensatory measures are performed or

diverse instruments continue to be available.

[[Page 25107]]

7. DEC proposed to change Section 4.5.4.b, which currently requires

verification of the refueling water storage tank temperature to be

within the allowed range once per 24 hours if the outside air

temperature is less than 70 degrees or greater than 100 degrees

Fahrenheit. The proposed change would simply require that the tank

temperature be verified within range every 24 hours regardless of

outside air temperature.

8. DEC proposed to revise Table 3.7-1, which imposes limits on the

maximum allowable power range neutron flux high setpoint for various

numbers of inoperable safety valves on any operating steam generator.

The revision would reduce the setpoints, making them more conservative.

9. Section 3.7.6, regarding the condensate storage system,

currently only exists in the Unit 2 TS. DEC proposed to impose these

requirements also on Unit 1.

10. Several electrical busses and inverters currently covered by

Section 3.8.3.1 are qualified by a footnote, which specifies the

conditions under which the inverter may be disconnected from its direct

current source. DEC proposed to delete this footnote because it is not

needed.

Basis for proposed no significant hazards consideration

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

provided its analyses of the issue of no significant hazards

consideration for each of the above proposed changes. The NRC staff has

reviewed the licensee's analyses against the standards of 10 CFR

50.92(c). The NRC staff's analysis is presented below.

1. Will the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

For changes 1, 2, 4, 5, 6, 7, 8, 9, and 10, the answer is ``no.''

The proposed changes will not affect the safety function of the subject

systems. There will be no direct effect on the design or operation of

any plant structures, systems, or components. No previously analyzed

accidents were initiated by the functions of these systems, and the

systems were not factors in the consequences of previously analyzed

accidents. Therefore, the proposed changes will have no impact on the

consequences or probabilities of any previously evaluated accidents.

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

kind of accident from any accident previously evaluated?

For changes 1, 2, 4, 5, 6, 7, 8, 9, and 10, the answer is ``no.''

The proposed changes would not lead to any hardware or operating

procedure change. Hence, no new equipment failure modes or accidents

from those previously evaluated will be created.

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

safety?

For changes 1, 2, 4, 5, 6, 7, 8, 9, and 10, the answer is ``no.''

Margin of safety is associated with confidence in the design and

operation of the plant. The proposed changes to the TS do not involve

any change to plant design, operation, or analysis. Thus, the margin of

safety previously analyzed and evaluated is maintained.

Based on this analysis, it appears that the three standards of 10

CFR 50.92(c) are satisfied for each of the proposed changes. Therefore,

the NRC staff proposes to determine that the amendment request involves

no significant hazards consideration.

Local Public Document Room location: York County Library, 138 East

Black Street, Rock Hill, South Carolina

Attorney for licensee: Mr. Paul R. Newton, Legal Department (PB05E),

Duke Energy Corporation, 422 South Church Street, Charlotte, North

Carolina

NRC Project Director: Herbert N. Berkow

Duke Energy Corporation (DEC), et al., Docket Nos. 50-413 and 50-414,

Catawba Nuclear Station, Units 1 and 2, York County, South Carolina

Date of amendment request: April 8, 1998.

Description of amendment request: The proposed amendments would

revise Section 3.6.5.1 and 4.6.5.1 of the Technical Specifications (TS)

of each unit to relax ice condenser stored ice weight requirements by

approximately 6 percent. The proposed change is based mainly on DEC's

gathered data showing lower sublimation rate than originally

anticipated.

Basis for proposed no significant hazards consideration

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

provided its analyses of the issue of no significant hazards

consideration for the proposed changes. The NRC staff has reviewed the

licensee's analyses against the standards of 10 CFR 50.92(c). The NRC

staff's analysis is presented below.

1. Will the changes involve a significant increase in the

probability or consequences of an accident previously evaluated?

No. The proposed changes will not affect the safety function of the

ice condenser in that there will be no changes to the design or

operation of any plant structures, systems, or components. No

previously analyzed accidents were initiated by the functions of the

ice condenser, and the ice condenser will remain fully capable of

performing its design accident mitigation function. Therefore, the

proposed changes will have no impact on the consequences or

probabilities of any previously evaluated accidents.

2. Will the changes create the possibility of a new or difference

kind of accident from any accident previously evaluated?

No. The proposed changes would not lead to any hardware or

operating procedure change. Reducing the required ice weight will not

have any impact on other plant systems that were assumed to be accident

initiators. Hence, no new equipment failure modes or accidents from

those previously evaluated will be created.

3. Will the changes involve a significant reduction in a margin of

safety? No. Margin of safety is associated with confidence in the

design and operation of the plant; specifically, the ability of the

fission product barriers to perform their design functions during and

following an accident. The proposed changes regarding required ice

weight do not involve any change to plant design, operation, or

analysis. Thus, the margin of safety previously analyzed and evaluated

is maintained.

Based on this analysis, it appears that the three standards of 10

CFR 50.92(c) are satisfied for the proposed changes. Therefore, the NRC

staff proposes to determine that the amendment request involves no

significant hazards consideration.

Local Public Document Room location: York County Library, 138 East

Black Street, Rock Hill, South Carolina

Attorney for licensee: Mr. Paul R. Newton, Legal Department (PB05E),

Duke Energy Corporation, 422 South Church Street, Charlotte, North

Carolina

NRC Project Director: Herbert N. Berkow

Duke Energy Corporation (DEC), Docket Nos. 50-369 and 50-370, McGuire

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

Date of amendment request: May 27, 1997.

Description of amendment request: The proposed changes would lower

the minimum required diesel generator (DG) air start receiver pressure

from 220 per square inch gauge (psig) to 210 psig with a monthly

verification, and would include an allowed outage time of 48 hours for

a degraded air receiver provided the redundant air receiver is

maintained at equal to or greater than 210 psig. These proposed changes

are associated with DEC's application to convert to the Improved

Technical

[[Page 25108]]

Specifications. Also, they are considered less restrictive requirements

because of the lower required minimum pressure and the allowance of

continued operation with a degraded starting air system.

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 for each change, which is presented below:

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

probability or consequence of an accident previously evaluated?

The proposed changes provide Actions for degraded capabilities

of the diesel starting air subsystems for the DG. The proposed

Actions establish limits for the DG starting air subsystems of 210

psig, (are) allowed to decrease below the required value for 48

hours(, and are verified every 31 days.) The Completion Times are

based on the amount of capability remaining, and the time needed to

correct any deficient condition. If the Completion Times are

exceeded, the specification requires the associated DG to be

declared inoperable immediately, consistent with the current TS

(technical specifications). Since the new Actions continue to assure

that the associated DG remains capable of performing its design

safety function, the proposed (changes do) not significantly affect

the probability or consequences of an accident previously evaluated.

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

kind of accident from any accident previously evaluated?

The proposed (changes do) not permit operation in a new or

different mode, or permit the installation of a new or different

type of equipment. The proposed changes provide Actions for degraded

capabilities of the DG starting air subsystems. The proposed Actions

establish Conditions, Required Actions, and Completion Times to be

entered when in a degraded condition. The DG remains capable of

performing its design safety function. Therefore, the proposed

(changes do) not create the possibility of a new or different kind

of accident from those previously evaluated.

3. (Do these changes) involve a significant reduction in a

margin of safety?

The proposed (changes do) not significantly increase the

probability or consequences of an accident previously evaluated. The

changes provide assurance that timely action will be initiated to

restore DG starting air subsystem when inoperabilities exist,

without unnecessarily forcing plant shutdown. Based on the limit for

the starting air subsystem for the DG, the limited time allowed is

acceptable to restore the parameter to within the requirements

without unnecessary plant shutdown. Therefore, (these 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: J. Murrey Atkins Library,

University of North Carolina at Charlotte, 9201 University City

Boulevard, Charlotte, North Carolina

Attorney for licensee: Mr. Albert Carr, Duke Energy Corporation, 422

South Church Street, Charlotte, North Carolina

NRC Project Director: Herbert N. Berkow

Duke Energy Corporation (DEC), Docket Nos. 50-369 and 50-370, McGuire

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

Date of amendment request: May 27, 1997.

Description of amendment request: The two proposed changes are

associated with DEC's application to convert to the Improved Technical

Specifications and are considered as administrative changes. The first

change would delete a current requirement to only verify the refueling

water storage tank temperature once every 24 hours if the outside air

temperature is less than 70 degrees or greater than 100 degrees

Fahrenheit, and would require that the tank temperature be verified

within range every 24 hours regardless of the outside air temperature

value. The second change would delete the current requirement that 32

of 33 hydrogen igniters be operable on each train, and would require

that 34 igniters per train to be operable. The actual design contains

35 igniters per train. This change would correct an inadvertent error

in the current Technical Specifications (TS). The number of igniters

was increased to 35 after the first refueling outage of each unit. This

change would correct the TS to reflect the requirements stated in

Safety Evaluation Report Supplement 7.

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 for each of the above proposed changes. The NRC staff has

reviewed the licensee's analyses against the standards of 10 CFR

50.92(c). The NRC staff's analysis is presented below:

1. Will the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed changes will not affect the safety function of the

subject systems. There will be no direct effect on the design or

operation of any plant structures, systems, or components. No

previously analyzed accidents were initiated by the functions of these

systems, and the systems were not factors in the consequences of

previously analyzed accidents. Therefore, the proposed changes will

have no impact on the consequences or probabilities of any previously

evaluated accidents.

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

kind of accident from any accident previously evaluated?

The proposed changes would not lead to any hardware or operating

procedure change. Hence, no new equipment failure modes or accidents

from those previously evaluated will be created.

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

safety?

Margin of safety is associated with confidence in the design and

operation of the plant. The proposed changes to the TS do not involve

any change to plant design, operation, or analysis. Thus, the margin of

safety previously analyzed and evaluated is maintained.

Based on this analysis, it appears that the three standards of 10

CFR 50.92(c) are satisfied for each of the proposed changes. Therefore,

the NRC staff proposes to determine that the amendment request involves

no significant hazards consideration.

Local Public Document Room location: J. Murrey Atkins Library,

University of North Carolina at Charlotte, 9201 University City

Boulevard, North Carolina

Attorney for licensee: Mr. Albert Carr, Duke Energy Corporation, 422

South Church Street, Charlotte, North Carolina

NRC Project Director: Herbert N. Berkow

Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear

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

Date of amendment request: May 27, 1997.

Description of amendment request: The proposed change would allow

two charging pumps or safety injection pumps capable of injecting into

the Reactor Coolant System (RCS) when the RCS is depressurized and an

RCS vent of at least 4.5 square inches is established. This proposed

change is associated with the licensee's application to convert to the

Improved Technical Specifications and results in a requirement less

restrictive than the current requirement.

Basis for proposed no significant hazards consideration

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

provided its analysis of the

[[Page 25109]]

issue of no significant hazards consideration for each change, which is

presented below:

1. Does the change involve a significant increase in the

probability or consequence of an accident previously evaluated?

The proposed change will provide an additional alternative for

low temperature (overpressure) relief capacity when two charging

pumps or safety injection pumps are capable of injecting into the

RCS. The low temperature (overpressure) protection is not considered

to be an initiator of any analyzed event, therefore, the proposed

change does not increase the probability of a previously analyzed

event.

The proposed change provides an equivalent vent size to the

existing two open PORVs (power-operated relief valves). Therefore,

this 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?

The proposed change does not necessitate a physical alteration

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

installed) or changes in the manner in which the plant is operated.

The proposed change adds an additional alternative to overpressure

protection equivalent to the current requirements. Therefore, the

proposed change will not create the possibility of a new or

different kind of accident than any previously evaluated.

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

of safety?

As described above, the proposed change adds an additional

alternative to overpressure protection equivalent to the current

requirements. The inclusion of additional alternatives provides the

operating staff with additional flexibility in meeting low

temperature overpressure protection requirements. Therefore, the

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: J. Murrey Atkins Library,

University of North Carolina at Charlotte, 9201 University City

Boulevard, Charlotte, North Carolina

Attorney for licensee: Mr. Albert Carr, Duke Energy Corporation, 422

South Church Street, Charlotte, North Carolina

NRC Project Director: Herbert N. Berkow

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

Station, Unit 3, St. Charles Parish, Louisiana

Date of amendment request: March 25, 1998

Description of amendment request: Revise Technical Specification

(TS) 3.9.8.1, ``Shutdown Coolant and Coolant Circulation High Water

Level,'' and TS 3.9.8.2, ``Shutdown Cooling and Coolant Circulation Low

Water Level,'' to change the minimum water level above the fuel

assemblies seated in the reactor vessel at which the Shutdown Cooling

(SDC) System is required to be maintained operable, or be in operation.

In addition, TS 3.8.1.2, ``Electric Power Systems, A.C. Sources,

Shutdown,'' and Technical Specification Bases 3/4.9.8, ``Shutdown

Cooling and Coolant Circulation,'' have been changed to make the

wording consistent with TS 3.9.8.1 and TS 3.9.8.2.

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

consequence of any accident?

Response: No.

The operation of the facility in accordance with this change

does not involve an increase in the probability of any accident.

Changing the water level at which the Shutdown Cooling (SDC)

System is required to be maintained operable or be in operation will

not increase the probability or consequences of an accident. The

design, operation, or configuration of the SDC system will not be

changed.

At least one shutdown cooling train will be in operation to

ensure sufficient cooling capacity is available to remove decay heat

and maintain the water in the reactor pressure vessel below 140

degree F as required during the refueling mode.

At least one shutdown cooling train will be in operation to

ensure sufficient coolant circulation is maintained through the

reactor core to minimize the effects of a boron dilution incident

and prevent boron stratification. Technical Specification 3.9.10.1,

``Refueling Operations Water Level--Reactor Vessel Fuel

Assemblies,'' will be complied with, and therefore, the assumptions

related to iodine removal and the fuel handling accident will be

preserved.

Sufficient time, approximately 1.00 hours, will be available to

the operators to initiate compensatory measures to preclude the

initiation of core boiling in the unlikely event SDC should be loss

[lost].

2. Will operation of the facility in accordance with this

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

accident from any accident previously evaluated?

Response: No.

The operation of the facility in accordance with this proposed

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

accident from any accident previously evaluated.

The proposed change will not affect the design, configuration,

or operation of the SDC system, and therefore there are no new modes

of failure introduced.

3. Will operation of the facility in accordance with this

proposed change involve a significant reduction in a margin of

safety?

Response: No.

Operation of the facility in accordance with this proposed

change will not involve a significant reduction in a margin of

safety.

The calculation of the time to the initiation of boiling based

on 23 feet above the top of the fuel seated in the reactor vessel,

at four days after shutdown, demonstrates there is significant time

available, approximately 1.00 hour, to the operators within which to

take compensatory measures to preclude the initiation of boiling.

The calculation shows that based on 23 feet of water above the

reactor flange there is 2.04 hours to the initiation of boiling.

Although there is a reduction in the time to the initiation of

boiling, compensatory measures could be taken within a few minutes

to restore SDC, and thus, there is still a significant margin

available to the operators within which to preclude the initiation

of boiling. Thus, the margin of safety is not significantly reduced.

The time to core uncovery was determined to be 27.74 hours based

on four days after shutdown and water level twenty-three (23) feet

above the fuel assemblies seated in the reactor vessel.

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 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 DC 20005-3502

NRC Project Director: John N. Hannon

Florida Power Corporation, et al., Docket No. 50-302, Crystal River

Nuclear Generating Plant, Unit No. 3, Citrus County, Florida

Date of amendment request: March 20, 1998.

Description of amendment request: The proposed amendment requests

editorial changes to the Improved Technical Specifications (ITS) Safety

Limits and Administrative Controls to replace the titles of the Senior

Vice President, Nuclear Operations (SVPNO) and the Vice President,

Nuclear Production (VPNP) with the position of Chief Nuclear Officer

(CNO). The CNO combines the duties of the SVPNO and VPNP as currently

described in ITS and is required to be an officer of the company. The

proposed change is

[[Page 25110]]

intended to allow upgrading the position of the corporate officer

responsible for overall nuclear operations without limiting the title.

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.

Does not involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed amendment does not involve a significant increase

in the probability or consequences of an accident previously

evaluated because the deletion and updating of individual titles

does not affect plant operation. No design basis accidents are

affected by the proposed administrative and editorial changes and,

as such, there are no physical changes to the facility or its

operation.

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

accident from any accident previously evaluated.

The proposed ITS changes are administrative and editorial in

nature. No changes to the facility structures, systems and

components or their operation will result. The design and design

basis of the facility remain unchanged. The plant safety analyses

remain current and accurate. No new or different failure mechanisms

are introduced. Therefore, the possibility of a new or different

kind of accident from any accident previously evaluated is not

introduced.

Does not involve a significant reduction in the margin of

safety.

The proposed ITS changes are administrative and editorial in

nature. The proposed safety margins established through the design

and facility license including the Improved Technical Specifications

remain unchanged. In addition, the proposed amendment ensures

continued emphasis and assignment of responsibility for overall

nuclear safety. Therefore, all margins of safety are maintained.

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

this review, it appears that the three standards of Sec. 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: Coastal Region Library, 8619 W.

Crystal Street, Crystal River, Florida 34428

Attorney for licensee: R. Alexander Glenn, General Counsel, Florida

Power Corporation, MAC-A5A, P.O. Box 14042, St. Petersburg, Florida

33733-4042

NRC Project Director: Frederick J. Hebdon

Florida Power Corporation, et al., Docket No. 50-302, Crystal River

Nuclear Generating Plant, Unit No. 3, Citrus County, Florida

Date of amendment request: March 20, 1998.

Description of amendment request: The proposed amendment would

change the Inservice Inspection Program described in Improved Technical

Specification (ITS) 5.6.2.8.c. This ITS currently states that the

reactor coolant pump (RCP) motor flywheels will be inspected during the

``Spring 1998 refueling outage,'' which would have been refueling

outage 11. Due to a recent 17-month extended outage, refueling outage

11 has been deferred until Fall 1999. The proposed change is intended

to accurately reflect the new refueling outage 11 schedule.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below.

The proposed change will not significantly increase the

probability or consequences of an accident previously evaluated.

The safety function of the RCP flywheels is to provide a

coastdown period during which the RCPs would continue to provide

reactor coolant flow to the reactor after loss of power to the RCPs.

The maximum loading on the RCP motor flywheel results from overspeed

following a large loss of coolant accident (LOCA). The estimated

maximum obtainable speed in the event of a Reactor Coolant System

piping break was established conservatively. The proposed one-time

editorial change to remove the words ``Spring 1998 refueling

outage'' and replace them with ``to coincide with Refueling Outage

11R'' does not affect that analysis. The proposed change in dates is

editorial in that it merely reflects the new date for cycle 11. The

usage time for the flywheels is bounded by the original estimates.

The proposed editorial change does not affect the amount of

radioactive material available for release or modify any systems

used for mitigation of such releases during accident conditions.

Therefore, the proposed change does not involve a significant

increase in the probability or consequences of any accident

previously evaluated.

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

different kind of accident from any accident previously evaluated.

The proposed editorial change will not change the design,

configuration, or method of operation of the plant. Therefore, the

proposed change will not create the possibility of a new or

different kind of accident from any previously evaluated.

The proposed change will not involve a significant reduction to

any margin of safety.

The proposed Amendment is an editorial change to reflect that

CR-3's operating cycle is not ending in spring 1998, but in fall

1999. The proposed change does not affect the methods of inspection

or its acceptance criteria. Therefore, the margins of safety defined

in RG [Regulatory Guide] 1.14 are not changed.

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

this review, it appears that the three standards of Sec. 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: Coastal Region Library, 8619 W.

Crystal Street, Crystal River, Florida 34428

Attorney for licensee: R. Alexander Glenn, General Counsel, Florida

Power Corporation, MAC-A5A, P.O. Box 14042, St. Petersburg, Florida

33733-4042

NRC Project Director: Frederick J. Hebdon

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn

County, Iowa

Date of amendment request: April 15, 1998.

Description of amendment request: The proposed amendment would

update the existing pressure-temperature curves with new curves with

values from 18 to 32 effective full power years based on the testing

and analysis of reactor pressure vessel surveillance materials.

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 pressure-temperature limits are not

derived from Design Basis Accident (DBA) analyses. They are

prescribed by the ASME B&PV Code and 10 CFR part 50 appendices G and

H as restrictions on normal operation to avoid encountering

pressure, temperature, and temperature rate of change conditions

that might cause undetected flaws to propagate and cause nonductile

failure of the reactor coolant pressure boundary.

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

new or different kind of accident from any accident previously

evaluated. The amendment will merely update the pressure-temperature

curves (and associated SRs and Bases) already existing in the plant

Improved Technical Specifications to provide limits from 18 to 32

EFPY of operation, which are based upon evaluation and analysis of

actual in-vessel material specimens, per 10 CFR part

[[Page 25111]]

50, appendices G and H. The pressure-temperature curves are

established to the requirements of 10 CFR part 50, appendix G to

assure that brittle fracture of the reactor vessel is prevented.

(3) The proposed amendment will not involve a significant

reduction in a margin of safety. 10 CFR part 50, appendix G

specifies fracture toughness requirements to provide adequate

margins of safety during operation over the service lifetime. The

values of adjusted reference temperature and upper shelf energy

determined as a result of the 10 CFR part 50, appendices G and H

analysis are expected to remain within the limits of Regulatory

Guide 1.99, Revision 2 and appendix G of 10 CFR part 50 (less than

200 deg. F and greater than 50 ft-lbs respectively) for at least 32

EFPY of operation.

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: Cedar Rapids Public Library, 500

First Street, SE., Cedar Rapids, IA 52401

Attorney for licensee: Jack Newman, Al Gutterman, Morgan, Lewis &

Bockius, 1800 M Street, NW., Washington, DC 20036-5869

Acting NRC Project Director: Richard P. Savio

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear

Station, Nemaha County, Nebraska

Date of amendment request: March 27, 1997.

Description of amendment request: The proposed amendment, included

as part of the proposed conversion from the current Technical

Specifications (TS) to improved TS, would establish Allowable Values

for the instrumentation included in Section 3.3, as a result of the

plant-specific application of the General Electric Instrument Setpoint

Methodology to the Cooper Nuclear Station (CNS).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change in selected Allowable Values for the

instrumentation included in proposed Section 3.3 of the Technical

Specifications is the result of application of the CNS

instrumentation setpoint methodology. This methodology incorporates

the guidance of ISA Recommended Practice ISA-RP67.04, Part II,

``Methodologies for the Determination of Setpoints for Nuclear

Safety-Related Instrumentation,'' September 1994. Application of

this methodology results in instrumentation selected Allowable

Values which more accurately reflect total instrumentation loop

accuracy as well as that of test equipment and setpoint drift

between Surveillances. The proposed change will not result in any

hardware changes. The instrumentation included in proposed Section

3.3 of the Technical Specifications is not assumed to be an

initiator of any analyzed event. Existing operating margin between

plant conditions and actual plant setpoints is not significantly

reduced due to this change. As a result, the proposed change will

not result in unnecessary plant transients.

The role of the proposed Section 3.3 instrumentation is in

mitigating and thereby limiting the consequences of accidents. The

Allowable Values have been developed to ensure that the design and

safety analysis limits will be satisfied. The methodology used for

the development of the Allowable Values ensures the affected

instrumentation remains capable of mitigating design basis events as

described in the safety analyses and that the results and

consequences described in the safety analyses remain bounding.

Additionally, the proposed change does not alter the plant's ability

to detect and mitigate events. Therefore, this 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?

The proposed change is the result of application of the CNS

instrumentation setpoint methodology and do not create the

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

previously evaluated. This is based on the fact that the method and

manner of plant operation is unchanged. The use of the proposed

Allowable Values does not impact safe operation of CNS in that the

safety analysis limits will be maintained. The proposed Allowable

Values involve no system additions or physical modifications to

systems in the station.

These Allowable Values were developed using a methodology to

ensure the affected instrumentation remains capable of mitigating

accidents and transients. Plant equipment will not be operated in a

manner different from previous operation, except that setpoints may

be changed. Since operational methods remain unchanged and the

operating parameters have been evaluated to maintain the station

within existing design basis criteria, no different type of failure

or accident is created.

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

of safety?

The proposed change does not involve a reduction in a margin of

safety. The proposed changes have been developed using a methodology

to ensure safety analysis limits are not exceeded. As such, this

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: Auburn Memorial Library, 1810

Courthouse Avenue, Auburn, NE 68305

Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power

District, Post Office Box 499, Columbus, NE 68602-0499

NRC Project Director: John N. Hannon

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear

Station, Nemaha County, Nebraska

Date of amendment request: March 27, 1997.

Description of amendment request: The proposed amendment, included

as part of the proposed conversion from the current Technical

Specifications (CTS) to the improved Technical Specifications (ITS),

would add an additional action statement to a limiting condition for

operation (LCO). The LCO is in the Improved Standard Technical

Specifications (ISTS, NUREG-1433, Revision 1) 3.6.2.3 on the residual

heat removal suppression pool cooling subsystems. The requirements in

the proposed ITS 3.6.2.3 on the subsystems do not exist in the CTS. The

Action B for ITS 3.6.2.3 would require that if the two such subsystems

were inoperable, one subsystem would have to be restored to operability

within 8 hours or the plant would be in ITS 3.0.3. ITS 3.0.3 governs

plant operation if an LCO (i.e., ISTS 3.6.2.3) and the associated

action statement are not met (i.e., Action B to ISTS 3.6.2.3).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change provides more stringent requirements for

operation of the facility. These more stringent requirements do not

result in operation that will increase the probability of initiating

an analyzed event and do not alter assumptions relative to (the)

mitigation of an accident or transient event. The more restrictive

requirements continue to ensure * * * systems, and components

((i.e., the residual heat removal suppression pool cooling

subsystems)) are maintained consistent with the safety analyses and

licensing basis. Therefore, this (the proposed)

[[Page 25112]]

change does not involve a significant (an) increase in the

probability or consequences of any accident previously evaluated.

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

kind of accident from any accident previously evaluated?

The proposed change does not involve a physical alteration of

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

or changes in the methods governing normal plant operation. The

proposed change does impose different requirements. However, this

change is consistent with the assumptions in the safety analyses and

licensing basis. Thus, this change does not create the possibility

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

evaluated is not created.

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

of safety?

The imposition of more restrictive requirements either has no

impact on or increases the margin of plant safety. As provided in

the discussion of the change, each change in this category (i.e.,

more restrictive requirements) is, by definition, providing

additional restrictions to enhance plant safety. The change

maintains requirements (systems and components) within the safety

analyses and licensing basis. Therefore, this 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: Auburn Memorial Library, 1810

Courthouse Avenue, Auburn, NE 68305

Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power

District, Post Office Box 499, Columbus, NE 68602-0499

NRC Project Director: John N. Hannon

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear

Station, Nemaha County, Nebraska

Date of amendment request: March 27, 1997.

Description of amendment request: The proposed amendment, included

as part of the proposed conversion from the current Technical

Specifications (CTS) to the improved Technical Specifications (ITS),

would add an additional test (i.e., water and sediment content within

limits) of diesel fuel oil that could be used in place of a current

test (i.e., clear and bright appearance with proper color) in the

diesel fuel oil testing program. The current tests are listed in CTS

4.9.A.2.d/e. The testing program will be in the new ITS 5.5.9. The

additional test is change number 25 to Section 5.0 of the Improved

Standard Technical Specifications (NUREG-1433, Revision 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. Does the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

The proposed change provides more stringent requirements for

operation of the facility. (This) more stringent (requirement)

do(es) not result in operation that will increase the probability of

initiating an analyzed event and do(es) not alter assumptions

relative to (the) mitigation of an accident or transient event. The

more restrictive (requirement) continue(s) to ensure * * * systems

and components (i.e., the diesel generators) are maintained

consistent with the safety analyses and licensing basis. Therefore,

the proposed change does not involve an increase in the probability

or consequences of any accident previously evaluated.

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

kind of accident from any accident previously evaluated?

The proposed change does not involve a physical alteration of

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

or changes in the methods governing normal plant operation. However,

this change is consistent with the assumptions in the safety

analyses and licensing basis. Thus, this change does not create the

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

previously evaluated is not created.

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

of safety?

The imposition of more restrictive requirements either has no

impact on or increases the margin of plant safety. As provided in

the discussion of the change, each change in this category (i.e., a

more restrictive requirement) is, by definition, providing

additional restrictions to enhance plant safety. The change

maintains (systems and components) within the safety analyses and

licensing basis. Therefore, this 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: Auburn Memorial Library, 1810

Courthouse Avenue, Auburn, NE 68305

Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power

District, Post Office Box 499, Columbus, NE 68602-0499

NRC Project Director: John N. Hannon

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear

Station, Nemaha County, Nebraska

Date of amendment request: March 27, 1997.

Description of amendment request: The proposed amendment, included

as part of the proposed conversion from the current Technical

Specifications (TS) to improved TS for the Cooper Nuclear Station

(CNS), would relocate the Trip Level Settings for the Rod Block Monitor

from Table 3.2.C of the current TS to the Core Operating Limits Report.

Also, details relating to the Alternate Shutdown system design and

operation are proposed to be relocated from current TS 3.2.I and 4.2.I

to the improved TS Bases.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration. The NRC staff has reviewed the licensee's analysis

against the three criteria of 10 CFR 50.92(c), and has determined the

following:

The proposed changes relocate certain details from the Technical

Specifications to the Bases and the Core Operating Limits Report

(COLR). The Bases and the COLR containing the relocated information

will be maintained in accordance with 10 CFR 50.59. In addition, the

Bases and COLR are subject to the applicable change control provisions

of Chapter 5.0, Administrative Controls'', of the proposed improved

Technical Specifications. Since any changes to the Bases or the COLR

will be evaluated per the requirements of 10 CFR 50.59 or other

applicable change control provisions, no increase in the probability or

consequences of an accident previously evaluated will result.

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

probability or consequences of an accident previously evaluated.

The proposed changes do not involve any physical alterations to the

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

changes in the methods governing normal plant operation. The proposed

changes will not impose or eliminate any requirements, and adequate

control of the information will be maintained. Thus, these changes do

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

any accident previously evaluated.

The proposed changes will not reduce a margin of safety because

they have no impact on any safety analysis assumptions. In addition,

the details to be transposed from the TS to the Bases

[[Page 25113]]

and the COLR are unchanged. Since any future changes to these details

in the Bases or the COLR will be evaluated per the requirements of 10

CFR 50.59 or other applicable change control provisions, no reduction

in a margin of safety will result. As such, these proposed changes do

not involve a significant reduction in a margin of safety.

Based on the above discussion, 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: Auburn Memorial Library, 1810

Courthouse Avenue, Auburn, NE 68305

Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power

District, Post Office Box 499, Columbus, NE 68602-0499

NRC Project Director: John N. Hannon

North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook

Station, Unit No. 1, Rockingham County, New Hampshire

Date of amendment request: April 8, 1998.

Description of amendment request: The proposed change would revise

Technical Specifications (TSs) 4.4.5.3, Steam Generators--Inspection

Frequencies, and 3.4.6.2.c, Reactor Coolant System (RCS) Leakage, and

the associated bases to accommodate fuel cycles of up to 24 months with

respect to the allowed time interval between steam generator inservice

inspections.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of an accident previously evaluated.

Extending Surveillance Requirement (SR) 4.4.5.3 to accommodate a

24 month cycle for inspection of steam generator tubes structural

integrity, as well as, imposing a more restrictive Limiting

Condition for Operation (TS 3.4.6.2.c) for reactor coolant system

leakage through Category C-2 steam generators, will neither

exacerbate nor significantly increase the probability or

consequences of an accident previously evaluated in the Seabrook

Station [updated final safety analysis report] UFSAR.

The proposed changes to SR 4.4.5.3 do not alter the intent or

method by which the surveillances are conducted, do not involve

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

systems or components (SSCs) function, and do not modify the manner

in which the plant is operated.

The proposed change to TS 3.4.6.2.c imposes more restrictive

limits on plant operations due to RCS leakage through steam

generators. The proposed change does not involve physical changes to

the plant or alter the way a SSC functions.

The proposed changes to SR 4.4.5.3 and TS 3.4.6.2.c, and their

associated Bases, will not adversely affect the ability of the steam

generators to perform their intended safety function. Furthermore,

the proposed changes do not adversely affect the physical protective

boundaries of the plant. The proposed changes do not affect accident

initiators or precursors and do not alter the design assumptions,

conditions, configuration of the facility or the manner in which the

plant is operated. The proposed changes do not alter or prevent the

ability of SSCs to perform their intended function to mitigate the

consequences of an initiating event within the acceptance limits

assumed in the Updated Final Safety Analysis Report (UFSAR). The

proposed changes are administrative in nature and do not change the

level of programmatic controls or the procedural details associated

with aforementioned surveillance requirements. While the proposed

changes will lengthen the interval between surveillances, the

increase in interval has been evaluated; and based on the reviews of

the steam generator tube eddy current test (ECT) inspections, it is

concluded that the wear growth rate of the only active degradation

mechanism (Anti-Vibration Bar (AVB) wear) identified to date at

Seabrook Station is such that sufficient margin exists between the

plugging criteria and structural limit such that no tubes are

predicted to exceed the structural limit even with the longer

surveillance interval.

Since there are no changes to previous accident analyses, the

radiological consequences associated with these analyses remain

unchanged, therefore, the proposed changes do not involve a

significant increase in the probability or consequences of an

accident previously evaluated. Therefore, the proposed changes will

not significantly increase the probability or consequences of any

previously analyzed accident.

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

or different kind of accident from any previously analyzed.

The proposed changes to TS 3.4.6.2 and SR 4.4.5.3, and

associated Bases, do not alter the design assumptions, conditions,

configuration of the facility or the manner in which the plant is

operated. There are no changes to the source term, containment

isolation or radiological release assumptions used in evaluating the

radiological consequences in the Seabrook Station UFSAR. Existing

system and component redundancy is not being changed by the proposed

changes. The proposed changes have no impact on component or system

interactions. The proposed changes are administrative in nature and

do not change the level of programmatic controls and procedural

details associated with the aforementioned surveillance

requirements. Therefore, since there are no changes to the design

assumptions, conditions, configuration of the facility, or the

manner in which the plant is operated and surveilled, the proposed

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

accident from any previously analyzed.

3. The proposed changes do not involve a significant reduction

in a margin of safety.

The proposed change ( ) to the surveillance intervals for SR

4.4.5.3 is still consistent with the basis for the interval. The

intent or method of performing the surveillances remains unchanged.

The more restrictive limit for leakage through any one steam

generator placed in Category C-2, as well as, the requirement to do

an engineering assessment of steam generator tube integrity,

provides additional margin of ensuring safe plant operation.

In addition, there is no adverse affect on equipment design or

operation and there are no changes being made to the Technical

Specification required safety limits or safety system settings that

would adversely affect plant safety. The proposed changes are

administrative in nature and do not change the level of programmatic

controls and procedural details associated with the aforementioned

surveillance requirements. While the proposed changes will lengthen

the interval between surveillances, the increase in interval has

been evaluated; and based on the reviews of the steam generator tube

ECT inspections, it is concluded that the wear growth rate of the

only active degradation mechanism (AVB wear) identified to date at

Seabrook Station is such that sufficient margin exists between the

plugging criteria and structural limit such that no tubes are

predicted to exceed the structural limit even with the longer

surveillance interval. Therefore, extension of the current

surveillance intervals to accommodate a 24 month cycle will not

significantly degrade the ability, the availability or the

reliability of the steam generators to perform their intended safety

function, thus, it is concluded that there is no 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: Exeter Public Library, Founders

Park, Exeter, NH 03833

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

Northeast Utilities Service Company, PO Box 270, Hartford, CT 06141-

0270

NRC Project Director: Cecil O. Thomas

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

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

Date of amendment request: April 6, 1998.

Description of amendment request: The proposed amendment will

modify

[[Page 25114]]

the Technical Specifications (TSs) by (1) adding a surveillance

requirement to verify pressurizer heater capacity to TS 3.4.4,

``Reactor Coolant System--Pressurizer,'' (2) moving the identification

of the location of the containment air temperature detectors from the

surveillance requirements portion of TS 3.6.1.5, ``Containment

Systems--Air Temperature,'' to the TS Bases for Containment Systems,

Section 3/4.4.6.1.5, ``Air Temperature,'' and (3) modifying the action

statements and surveillance requirements of TS 3.7.1.5, ``Plant

Systems--Main Steam Isolation Valves.'' The TS Bases would also be

updated to include the list of containment air temperature detectors

and reflect the proposed 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. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change to add a surveillance requirement (SR)

4.4.4.2 to verify pressurizer heater capacity will help ensure the

pressurizer will be able to function as designed to maintain Reactor

Coolant System pressure. There will be no effect on any design basis

accident previously evaluated or on any equipment important to

safety. Therefore, the proposed change will not result in a

significant increase in the probability or consequences of an

accident previously evaluated.

The proposed changes to modify the wording of SR 4.6.1.5 and to

relocate the list of containment air temperature detectors from SR

4.6.1.5 to the Bases will not affect the Technical Specification

limit for containment temperature or the frequency of verification

of this limit. The proposed changes do not alter the way any

structure, system, or component functions. The initial assumption

for containment temperature used in the design basis accident

analysis will remain the same. There will be no affect on any design

basis accident previously evaluated or on any equipment important to

safety. Therefore, the proposed changes will not result in a

significant increase in the probability or consequences of an

accident previously evaluated.

The proposed changes to the action statements and surveillance

requirements of Technical Specification 3.7.1.5 will not affect the

operability requirements of the main (steamline) isolation valves

(MSIVs). There will be no effect on any design basis accident

previously evaluated or on any equipment important to safety.

Therefore, the proposed changes will not result in a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed changes have no adverse effect on any of the design

basis accidents previously evaluated or on any equipment important

to safety. Therefore, the License Amendment Request does not impact

the probability of an accident previously evaluated nor does it

involve a significant increase in the consequences or an accident

previously evaluated.

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

from any accident previously evaluated.

The proposed changes will not alter the plant configuration (no

new or different type of equipment will be installed) or require any

new or unusual operator actions. They do not alter the way any

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

manner in which the plant is operated. The proposed changes 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. Involve a significant reduction in a margin of safety.

The proposed changes will add SR 4.4.4.2 to verify pressurizer

heater capacity, relocate the list of containment temperature

detectors used to verify containment temperature from SR 4.6.1.5 to

the associated Bases, and modify the action statements and

surveillance requirements of Technical Specification 3.7.1.5.

These changes will have no adverse effect on equipment important

to safety. This equipment will continue to function as assumed in

the design basis accident analysis. Therefore, there will be no

significant reduction in the margin of safety as defined in the

Bases for the technical Specifications affected by these proposed

changes.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center, Three

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

Connecticut, and the Waterford Library, ATTN: Vince Juliano, 49 Rope

Ferry Road, Waterford, Connecticut

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

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

Connecticut

NRC Deputy Director: Phillip F. McKee

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

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

Date of amendment request: April 13, 1998

Description of amendment request: The proposed amendment would

change the Technical Specifications (TSs) by adding a new TS 3.5.5,

``Emergency Core Cooling Systems--Trisodium Phosphate (TSP).'' Also,

the surveillance requirements in TSs 4.5.2.c.3 and 4.5.2.c.4 would be

relocated to new TS 3.5.5 as TS 4.5.5.1 and TS 4.5.5.2, respectively.

The applicable TS Index page and Bases sections will be updated to

reflect the proposed changes.

Changes to the current requirements for the TSP are also proposed.

The TSP requirements in TS 4.5.2.c.3 would become the limiting

conditions for operation in the new TS; the amount of TSP required

would increase from ``equal to or greater than 110 cubic feet'' to

``equal to or greater than 282 cubic feet'' based on the new

calculations; the applicability would be expanded to include all of

Mode 3; the action statement would allow 48 hours to restore the TSP

volume; and changes would also be made to the required tests and

specific details would be relocated to the applicable TS Bases.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed changes to relocate the current trisodium phosphate

(TSP) dodecahydrate Technical Specification requirements from the

surveillance requirements for the Emergency Core Cooling System to a

new TSP Technical Specification will not change the requirement to

store TSP inside containment. The proposed changes will require a

large quantity of TSP to be stored inside containment. This large

quantity, based on a recently revised calculation, will ensure

sufficient TSP is available for containment sump water pH control.

These proposed changes do not alter the way any structure, system,

or component functions. There will be no adverse effect on any

design basis accident previously evaluated, on any equipment

important to safety, or o n the radiological consequences of any

design basis accident. Therefore, this License Amendment Request

does not impact the probability of an accident previously evaluated

nor does it involve a significant increase in the consequences of an

accident previously evaluated.

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

from any accident previously evaluated.

The proposed change to increase the TSP volume stored inside

containment will require two of the wire mesh TSP baskets inside

containment to be replaced by two new and larger wire mesh baskets.

The design of the new baskets has been evaluated and it is

consistent with the requirements for equipment installed in

containment. The replacement of the two wire mesh baskets

[[Page 25115]]

will not result in any significant change in plant configuration and

will not require any new or unusual operator actions. It will alter

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

alter the manner in which the plant is operated. It will 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. Involve a significant reduction in a margin of safety.

The proposed changes will relocate the current Technical

Specification requirements for TSP to a new Technical Specification.

The minimum required volume will be increased to reflect the results

of a new calculation performed to support the current requirement to

raise containment sump pH [equal to or greater than] 7.0. These

changes will have no adverse effect on equipment important to

safety. This equipment will continue to function as assumed in the

design basis accident analysis. Therefore, there will be no

significant reduction of the margin of safety as defined in the

Bases for the Technical Specifications affected by these proposed

changes.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center, Three

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

Connecticut, and the Waterford Library, ATTN: Vince Juliano, 49 Rope

Ferry Road, Waterford, Connecticut

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

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

Connecticut

NRC Deputy Director: Phillip F. McKee

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

Generating Plant, Wright County, Minnesota

Date of amendment request: April 11, 1997 (supersedes July 26,

1996, application)

Description of amendment request: The proposed amendment would

modify the Monticello Technical Specifications (TS) sections 3.6.C,

Coolant Chemistry, and 3/4.17.B, Control Room Emergency Filtration

System. The changes were proposed to establish TS requirements

consistent with modified analysis inputs used for the evaluation of the

radiological consequences of the main steam line break accident.

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.

A limit is established in the plant Technical Specifications for

steady state radioiodine concentration in the reactor coolant to

ensure that in the event of a release of radioactive material to the

environment due to a postulated high energy line break up to and

including a design basis Main Steam Line Break Accident, radiation

doses are maintained within the guidelines of 10 CFR part 100. The

steady state radioiodine concentration in the reactor coolant is an

input for analysis of the radiological consequences of an accident

due to a Main Steam Line Break outside of containment and postulated

high energy line breaks. In addition, requirements are established

in the Technical Specifications for control room habitability.

During an accident, the control room emergency filtration system

provides filtered air to pressurize the Control Room to minimize the

activity, and therefore the radiological dose, inside the control

room.

A change is proposed for the steady state radioiodine

concentration. This value is conservative with respect to the value

used in the Main Steam Line Break dose consequences analysis and is

consistent with the dose consequences evaluation of a postulated

Reactor Water Cleanup (RWCU) line break. Changes are proposed to the

limiting conditions for operation and surveillance requirements for

the Control Room Emergency Filtration Train iodine removal

efficiency. These changes are consistent with the inputs used in the

analysis of the radiological consequences of the postulated RWCU

line break and the Main Steam Line Break Accident. These proposed

requirements maintain operating restrictions for analytical inputs

used in the analysis of the Main Steam Line Break Accident.

Evaluation of these events has demonstrated that the postulated

radiological consequences will remain within the licensing basis

established in the AEC [Atomic Energy Commission] Provisional

Operating License Safety Evaluation Report, dated March 18, 1970,

thus the proposed changes do not result in an increase in the

consequences of previously evaluated accidents.

The analysis of the Main Steam Line Break Accident performed

using a reactor coolant radioiodine concentration of 2

(microcuries)/gm dose equivalent Iodine-131 and a control room

ventilation filter efficiency consistent with the proposed Technical

Specifications changes demonstrated that radiological consequences

of the Main Steam Line Break are not changed significantly. The

radiological consequences of the Main Steam Line Break Accident

remain within the exposure guidelines of 10 CFR part 100 and 10 CFR

part 50 appendix A, General Design Criterion 19. The offsite dose

consequences remain bounded by the licensing basis provided in the

AEC Provisional Operating License Safety Evaluation Report, dated

March 18, 1970. The control room doses calculated for the hot

standby Main Steam Line Break Accident using the TID-14844 dose

conversion factors remain bounded by the dose consequences of the

comparable design basis loss of coolant accident.

The evaluation of the postulated RWCU line break, performed

using a reactor coolant radioiodine concentration of 0.25

(microcurie)/gm dose equivalent Iodine-131 and a control room

ventilation filter efficiency consistent with the proposed Technical

Specifications changes, demonstrated that the radiological

consequences of this event remain within the exposure guidelines of

10 CFR part 100 and 10 CFR part 50 Appendix A, General Design

Criterion 19. The offsite dose consequences remain bounded by the

Main Steam Line Break as established in the licensing basis provided

in the AEC Provisional Operating License Safety Evaluation Report,

dated March 18, 1970.

The proposed Technical Specification changes do not introduce

new equipment operating modes, nor do the proposed changes alter

existing system inter-relationships. The proposed changes do not

introduce new failure modes. The system improvements to reduce

bypass leakage during postulated accidents do not have an adverse

effect on control room habitability. Therefore, this amendment will

not cause a significant increase in the probability of an accident

previously evaluated for the Monticello plant.

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

new or different kind of accident from any accident previously

analyzed.

The proposed Technical Specification changes do not introduce

new equipment operating modes, nor do the proposed changes alter

existing system inter-relationships. Operator action to mitigate the

consequences of the postulated RWCU line break is conservative based

on the very limited action required by the operator to close the

containment isolation valves and the availability of control room

indications to alert the operator to the postulated break. The use

of a ten (10) minute operator response time to take manual actions

in response to postulated events is consistent with Monticello's

licensing basis for similar events. The use of operator actions and

all available equipment is consistent with current regulatory

guidance for mitigating the consequences of postulated line breaks.

The proposed change to the specification for reactor coolant

dose equivalent radioiodine is conservative with respect to the re-

evaluation of the Main Steam Line Break Accident for the more

conservative hot standby initial condition for the postulated

accident. The proposed change to the specification for reactor

coolant dose equivalent radioiodine is consistent with the

postulated high energy line break of a Reactor Water Cleanup line.

The proposed changes to the limiting conditions for operation and

[[Page 25116]]

surveillance requirements for the control room emergency filtration

train iodine removal efficiency are consistent with the inputs used

in the evaluation of the radiological consequences of the postulated

RWCU line break and the Main Steam Line Break Accident. The system

improvements to reduce bypass leakage during postulated accidents do

not have an adverse effect on control room habitability. Therefore,

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

different kind of accident.

3. The proposed amendment will not involve a significant

reduction in the margin of safety.

Surveillance data has demonstrated the proposed requirements are

within the current capability of the facility. The proposed changes

maintain margins of safety. These proposed requirements maintain

operating restrictions for analytical inputs used in the analysis of

the bounding postulated high energy line break of a Reactor Water

Cleanup line and the Main Steam Line Break Accident. The proposed

change to the specification for reactor coolant dose equivalent

radioiodine is conservative with respect to the re-evaluation of the

Main Steam Line Break Accident for the more conservative hot standby

initial condition for the postulated accident. The proposed change

to the specification for reactor coolant dose equivalent radioiodine

is consistent with the postulated high energy line break of a

Reactor Water Cleanup line. The evaluation of these postulated

events determined that the radiological consequences remain within

the exposure guidelines of 10 CFR part 100 and of 10 CFR part 50

Appendix A, General Design Criterion 19. The proposed changes to the

limiting conditions for operation and surveillance requirements for

the control room emergency filtration train iodine removal

efficiency provide assurance that the system will perform at the

filter efficiency as used in the evaluation of the radiological

consequences of the postulated events. Therefore, the proposed

amendment will not involve a significant reduction in the margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401

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

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

NRC Project Director: Cynthia A. Carpenter

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: April 10, 1998.

Description of amendment request: The proposed amendments would

revise the combined Technical Specifications (TS) for the Diablo Canyon

Power Plant Unit Nos. 1 and 2 to revise TS 6.2.2.g and 6.3 to change

the name of the Operations Manager to Operations Director and to change

the requirement for the Operations Director to hold a senior reactor

operator (SRO) license.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed change to revise the title of the Operations

Manager to Operations Director is an administrative change that

clarifies the Technical Specification (TS) to reflect current

position titles.

The proposed change provides assurance that the Operations

Director will continue to have knowledge of pressurized water

reactor (PWR) operation and emergency event mitigation. The proposed

change does not detract from the Operations Director's ability to

perform his primary responsibilities. In this case, by having

previously held a senior reactor operator (SRO) license, the

Operations Director has achieved the necessary training, skills, and

experience to fully understand the operation of plant equipment and

the watch requirements for operators. In summary, the proposed

change does not affect the ability of the Operations Director to

provide the plant oversight required of his position.

Additionally, another off-shift individual that holds an SRO

license for Diablo Canyon Power Plant (DCPP) directs the licensed

activities of licensed operators (an Operations middle manager) will

have specific knowledge of operation and emergency event mitigation

at DCPP. This will assure that the change in qualification of the

Operations Director does not affect the probability of an operator

initiating an accident or increasing the consequences of an accident

due to improper direction from management. The training and

qualification programs for operators on shift will not be affected

by the proposed changes.

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 accident from any accident previously

evaluated.

The proposed change to revise the title of the Operations

Manager to Operations Director is an administrative change that

clarifies the TS to reflect current position titles.

The proposed change to TS 6.2.2g. and 6.3 do not affect the

design or function of any plant system, structure, or component, nor

does it change the way plant systems are operated. It does not

affect the performance of NRC licensed operators since the proposed

changes do not impact the training or qualification of any operator

on shift. Operation of the plant in conformance with TS and other

license requirements will continue to be supervised by personnel who

hold an SRO license. The proposed change to TS 6.2.2g and 6.3

ensures that the Operations Director will be a knowledgeable and

qualified individual by requiring the individual to have held an SRO

license at a PWR.

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 revise the title of the Operations

Manager to Operations Director is an administrative change that

clarifies the TS to reflect current position titles.

The proposed change involves an administrative control that is

not related to the margin of safety. The proposed change does not

reduce the level of knowledge or experience required of an

individual who fills the Operations Director position, nor does it

affect the conservative manner in which the plant is operated. The

on-shift licensed operators will continue to be supervised by

personnel who hold an SRO license in accordance with 10 CFR

50.54(l).

Therefore, neither of the proposed changes 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 Sec. 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 &

Electric Company, P.O. Box 7442, San Francisco, California 94120

NRC Project Director: William H. Bateman

[[Page 25117]]

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

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

Jersey

Date of amendment request: March 26, 1998.

Description of amendment request: The proposed amendments would

revise Technical Specification (TS) 3/4.8.2.1, ``AC Distribution--

Operating,'' to add operability conditions and action statements for

the 115-volt vital instrument bus (VIB) D and inverter. The proposed

amendments complete the recommended action from NRC Generic Letter 91-

11, Resolution of Generic Issues 48, ``LCOs for Class 1E Vital

Instrument Buses,'' and 49, ``Interlocks and LCOs for Class 1E Tie

Breakers'' pursuant to 10 CFR 50.54(f), dated July 18, 1991.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed change, as described above, does not make any

physical changes to the plant or components, nor changes the manner

in which the plant or components are operated as a result of the

addition of the Note and the D VIB and Inverter to the TS. The

proposed change incorporates the operating requirements of the

Technical Specification Interpretation (TSI) developed in response

to GL 91-11 into the Salem Unit 1 and 2 Technical Specifications.

Incorporating this interpretation into the Technical Specifications

eliminates the need for the TSI.

Therefore, the proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed change does not introduce any design or physical

configuration change to the plants, change the function of the 115

Volt D VIBs and inverters, or the manner in which they are

maintained or tested.

Therefore, the proposed amendment will not create the

possibility of a new or different kind of accident from any

previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed Action Times associated with the incorporation of

the D VIB into the Technical Specifications are consistent with the

current Action Times for the A, B, and C VIBs for a loss of an AC

bus. Adding the note to the Salem Unit 1 Technical Specification

brings consistency between Salem Units 1 and 2, and is also

consistent with NUREG 1431, Vol. 1, Rev 1 ``Standard Technical

Specifications Westinghouse Plants.''

The outage duration limit of 72 hours for the D inverter is

acceptable based on the following: (1) the proposed 72 hours Action

Time to restore the inoperable inverter to operable is supported by

a PSA [probabilistic safety assessment] assessment. NRC Draft SRP

[Standard Review Plan] Chapter 16.1, Revision 13, ``Risk-Informed

Decision making: Technical Specifications'' notes that an

incremental conditional core damage probability (ICCDP) of 5.0 E-7

is considered very small. The proposed 72 hour allowable outage time

was calculated utilizing the NRC incremental conditional core damage

probability (ICCDP), and (2) the inoperability of the D VIB Inverter

will not affect the operation of any Safeguard Equipment Cabinet

(SEC) or Emergency Diesel Generator (EDG).

Therefore, the proposed amendment 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: Salem Free Public Library, 112

West Broadway, Salem, NJ 08079

Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business

Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038

NRC Project Director: Robert A. Capra.

Toledo Edison Company, Centerior Service Company, and The Cleveland

Electric Illuminating Company, Docket No. 50-346, Davis-Besse Nuclear

Power Station, Unit 1, Ottawa County, Ohio

Date of amendment request: April 18, 1997, as supplemented by

letters dated October 10, 1997, and February 27, 1998.

Description of amendment request: The proposed amendment would

revise Technical Specification (TS) Section 3/4.7.6, ``Plant Systems--

Control Room Emergency Ventilation System.'' Additional Limiting

Conditions for Operation would be added related to the availability of

the station vent normal range radiation monitoring instrumentation. The

associated TS bases would also be modified consistent with these

changes. The staff's proposed no significant hazards consideration

determination for the requested change was published on June 4, 1997

(62 FR 30646).

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:

The Davis-Besse Nuclear Power Station has reviewed the proposed

changes and determined that a significant hazards consideration does

not exist because operation of the Davis-Besse Nuclear Power Station

(DBNPS), Unit No. 1, in accordance with this change would:

1a. Not involve a significant increase in the probability of an

accident previously evaluated because no accident initiators,

conditions, or assumptions are affected by the proposed changes.

The proposed change to Limiting Condition for Operation (LCO)

3.7.6.1 would include new required Action statements in the event

that one or both channels of Station Vent Normal Range Radiation

Monitoring instrumentation become inoperable. Under the proposed

Action statements for inoperable Station Vent Normal Range Radiation

Monitoring instrumentation, should the control room normal

ventilation system be isolated and at least one train of the control

room emergency ventilation system be placed in operation, these

systems would be in a state equivalent to that which they would be

in following an actual high radiation condition. These proposed

changes have no bearing on the probability of an accident.

The proposed change to the terminology utilized in Surveillance

Requirement (SR) 4.7.6.1.e is an administrative change made to make

the terminology consistent with the proposed new Action statements.

The proposed changes to Bases 3/4.7.6 are administrative changes

consistent with the proposed changes to LCO 3.7.6.1. These changes

have no bearing on the probability of an accident.

Not involve a significant increase in the consequences of an

accident previously evaluated because the proposed changes do not

change the source term, containment isolation, or allowable

releases.

As described above, under the proposed Action statements for

inoperable Station Vent Normal Range Radiation Monitoring

instrumentation, should the control room normal ventilation system

be isolated and at least one train of the control room emergency

ventilation system be placed in operation, these systems would be in

a state equivalent to that which they would be in following an

actual high radiation condition. Therefore, in the unlikely event of

an accident requiring control room isolation while in this

condition, the dose consequences to control room operators would be

unchanged.

The proposed change to the terminology utilized in Surveillance

Requirement (SR) 4.7.6.1.e is an administrative change made to make

the terminology consistent with the proposed new Action statements.

The proposed changes to Bases 3/4.7.6 are administrative changes

consistent with the proposed changes to LCO 3.7.6.1. These changes

have no bearing on the consequences of an accident.

2. Not create the possibility of a new or different kind of

accident from any accident

[[Page 25118]]

previously evaluated because no new accident initiators or

assumptions are introduced by the proposed changes.

As described above, under the proposed Action statements for

inoperable Station Vent Normal Range Radiation Monitoring

instrumentation, should the control room normal ventilation system

be isolated and at least one train of the control room emergency

ventilation system be placed in operation, these systems would be in

a state equivalent to that which they would be in following an

actual high radiation condition. Operation of the equipment and

components in this manner would not introduce the possibility of any

new or different kinds of accidents.

The proposed change to the terminology utilized in Surveillance

Requirement (SR) 4.7.6.1.e is an administrative change made to make

the terminology consistent with the proposed new Action statements.

The proposed changes to Bases 3/4.7.6 are administrative changes

consistent with the proposed changes to LCO 3.7.6.1. These changes

would not introduce the possibility of any new or different kinds of

accidents.

3. Not involve a significant reduction in a margin of safety

because the proposed changes to the Action under LCO 3.7.6.1 ensure

that control room isolation capability is maintained in the event a

station vent radiation monitor is inoperable. The proposed allowable

outage time of seven days for one inoperable channel is consistent

with the presently allowable outage time for one inoperable CREVS.

The proposed Action to place at least one CREVS train in operation

within one hour, in the event both channels of radiation monitoring

become inoperable, is more conservative than the present Action

which requires that a plant shutdown commence within one hour, but

does not require the CREVS be placed in operation.

The proposed change to the terminology utilized in Surveillance

Requirement (SR) 4.7.6.1.e is an administrative change made to make

the terminology consistent with the proposed new Action statements.

The proposed changes to Bases 3/4.7.6 are administrative changes

consistent with the proposed changes to LCO 3.7.6.1. These changes

would not affect the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of Toledo, William

Carlson Library, Government Documents Collection, 2801 West Bancroft

Avenue, Toledo, OH 43606

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

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

NRC Acting Project Director: Richard P. Savio

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,

Callaway County, Missouri

Date of application request: March 9, 1998.

Description of amendment request: The proposed amendment

application would revise Technical Specification 3/4.5.2b.1 and its

associated Bases to add clarification in regard to venting the

emergency core cooling system (ECCS) pump casings and accessible

discharge piping high points. Technical Specification 3/4.5.2b.1

requires verification that the ECCS piping is full of water at least

once per 31 days by venting the ECCS pump casings, i.e., the safety

injection pump, residual heat removal pump, and centrifugal charging

pump casings and accessible discharge piping high points. The

centrifugal charging pump (CCP) casings do not have installed casing

vents. Instead of a casing vent, the suction and discharge piping is

installed as vertical runs attached to the top-mounted suction and

discharge nozzles of each CCP pump. Information provided by the pump

manufacturer indicates that the vertical configuration of the piping is

sufficient to prevent the accumulation of noncondensible gases that

could cause gas binding. Therefore the CCP casings are effectively

vented by vents on the CCP discharge lines. The proposed amendment

application would revise Technical Specification 3/4.5.2b.1 and

associated Bases to require the residual heat removal and safety

injection pump casings and accessible ECCS discharge piping high points

be vented to ensure the ECCS piping is full of water.

Basis for proposed no significant hazards consideration

determination:

As required by 10 CFR 50.91(a), the licensee has provided its

analysis of the issue of no significant hazards consideration, which is

presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed change will align the surveillance requirements

with the installed system design and normal operating conditions.

The performance of surveillances required by Technical

Specifications is not postulated to initiate an accident. The intent

of the surveillance ensures OPERABILITY of the ECCS by verifying

that the ECCS piping is full of water and not subjected to gas

binding or water hammer. The design of the CCPs is such that

significant noncondensible gases do not collect in the pumps,

whether they are running or not. Therefore, it is unnecessary to

require periodic pump casing venting to ensure the CCPs will remain

OPERABLE. In addition, operating experience has shown that no

significant voiding has occurred in the affected piping which will

continue to be vented at a high point every 31 days per Surveillance

Requirement 4.5.2b.1). Therefore, no increase in the probability or

consequences of an accident will occur as a result of this change.

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

or different kind of accident from any accident previously

evaluated.

The proposed change will not result in new failure modes because

there are no hardware changes nor are there any changes in the

method by which any safety-related plant system performs its safety

function. The design of the CCPs is such that significant

noncondensible gases do not collect in the pumps, whether they are

running or not. Therefore, it is not necessary to require periodic

pump casing venting to ensure the equipment will remain OPERABLE.

Manual venting operations will be performed to minimize the

potential for voids in system piping. Accordingly, this change will

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

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed change does not affect the acceptance criteria for

any analyzed event. There will be no effect on the manner in which

safety limits or limiting safety system settings are determined nor

will there be any effect on those plant systems necessary to assure

the accomplishment of protective functions. There will be no impact

on any 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: University of Missouri-Columbia,

Elmer Ellis Library, Columbia, Missouri 65201-5149

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

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

NRC Project Director: William H. Bateman

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,

Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

Date of amendment request: December 18, 1997.

Description of amendment request: The proposed changes revise the

Technical Specifications (TS) to clarify the terminology used to

describe equipment surveillances performed with a refueling interval

frequency. Currently the TS are somewhat ambiguous in the wording in

this regard, and the proposed changes would adhere to the improved

Standard TS

[[Page 25119]]

and make it clear whether the reactor must be shutdown when performing

the test, or whether a ``refueling interval'' frequency (e.g., 18

months) is intended. All of the clarifications are in Section 4 of the

TS. In addition, minor typographical errors are being corrected, and an

obsolete reference is proposed to be deleted.

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:

Criterion 1--Operation of Surry Units 1 and 2 in accordance with

the proposed Technical Specifications change does not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

The probability of an accident is not increased as a result of

the proposed Technical Specification change since surveillance

intervals are being clarified, not changed, and will continue to

validate system/component availability, operability and performance

during the appropriate unit mode. The proposed change is

administrative in nature, therefore, station operations are not

being affected. The consequences of an accident previously evaluated

are not increased since station operations are not being changed,

and no physical modifications are being made to plant systems or

components.

Criterion 2--The proposed Technical Specifications change does

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

from any accident previously evaluated.

As noted above, the proposed change is administrative in nature.

A new or different type of accident is not being created since no

new accident precursors are being introduced and equipment

surveillances will continue to be performed as required to ensure

proper system/component operation. Plant systems are not being

modified, system operations are not being affected, and equipment

surveillance intervals are not being increased. Consequently, the

proposed change does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

Criterion 3--The proposed Technical Specifications change does

not involve a significant reduction in a margin of safety.

This is an administrative change. Clarification of refueling

surveillance interval terminology to ensure consistency in

application does not affect plant equipment performance.

Surveillance intervals are not being increased, and equipment

surveillance tests performed on a refueling interval frequency (i.e.

once per 18 months) will continue to ensure system/component

performance as assumed in the existing safety analyses. Therefore,

the proposed Technical Specification change does not involve a

reduction in the margin of safety.

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

this review, it appears that the three standards of Sec. 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: Swem Library, College of William

and Mary, Williamsburg, Virginia 23185

Attorney for licensee: Michael W. Maupin, Esq., Hunton and Williams,

Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, Virginia

23219

NRC Project Director: P.T. Kuo, Acting

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,

Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

Date of amendment request: March 25, 1998.

Description of amendment request: The proposed amendments would

revise the Technical Specifications (TS) Sections 6.1.A; 6.1.A.2;

6.1.C.1.a and b; 6.1.C.1.f.1,4 and 8; 6.1.C.1.g.1 and 3; 6.8.A.2; and

6.8.B.2 for Units 1 and 2, changing the title of Station Manager to

Site Vice President, and the titles of the Assistant Station Managers

to Manager-Station Operations and Maintenance and Manager-Station

Safety and Licensing.

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:

Virginia Electric and Power Company has reviewed the proposed

Technical Specifications changes against the criteria of 10 CFR

50.92 and has concluded that the changes do not pose a significant

hazards consideration. Specifically, station operations in

accordance with the proposed Technical Specifications changes will

not:

a. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed changes are administrative in nature. The overall

responsibility for safe operation and review of plant operations is

not being changed. There are no changes to the operation of any

plant system or its design as a result of these changes. Therefore,

neither the probability of occurrence nor the consequences of an

accident or malfunction of equipment important to safety previously

evaluated in the safety analysis report are increased.

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

from any accident previously evaluated.

The proposed changes are administrative in nature. The overall

responsibility for safe operation and review of plant operations is

not being changed. There are no changes to the operation of any

plant system or its design that could create any new modes of

operation or accident precursors. Therefore, it is concluded that no

new or different kind of accident or malfunction from any previously

evaluated has been created.

c. The proposed changes do not result in a significant reduction

in margin of safety as defined in the basis for any Technical

Specifications.

The proposed changes are administrative in nature. The overall

responsibility for safe operation and review is not being changed.

There are no changes to the operation of any plant system or its

design as a result of these changes. Safety systems are maintained

operable as required by Technical Specifications. Therefore, the

margin of safety is not changed.

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: Swem Library, College of William

and Mary, Williamsburg, Virginia 23185

Attorney for licensee: Michael W. Maupin, Esq., Hunton and Williams,

Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, Virginia

23219

NRC Project Director: P.T. Kuo, Acting

Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee

Nuclear Power Plant, Kewaunee County, Wisconsin

Date of amendment request: April 8, 1998.

Description of amendment request: The change would reduce allowable

reactor coolant system (RCS) specific activity from 1.0 microcurie/gram

to 0.35 microcurie/gram dose equivalent I-131.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The proposed change was reviewed in accordance with the

provisions of 10 CFR 50.92 to show no significant hazards exist. The

proposed change will not:

(1) Involve a significant increase in the probability or

consequence of an accident previously evaluated.

The change implements a more restrictive RCS activity limit.

Specific RCS activity is an initial plant condition and, therefore,

is not an accident initiator and can not cause the occurrence of or

increase the probability of an accident. The change also lowers the

curve of Figure TS 3.1-3 which restricts operation with high

specific activity. The new value for specific activity is justified

by

[[Page 25120]]

the Westinghouse calculation which demonstrates acceptable offsite

and control room doses following a (main steamline break) MSLB with

a maximum allowable primary to secondary leak rate. By lowering the

RCS specific activity and maintaining leakage within the projected

maximum allowable, 10 CFR 100 and GDC 19 criteria are satisfied.

Therefore, the change does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

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

accident from any previously evaluated.

The proposed change to the RCS specific activity limit will not

significantly effect operation of the plant nor will it alter the

configuration of the plant. There will be no additional challenges

to the main steam system or the reactor coolant system pressure

boundary and no new failure modes are introduced. Therefore, the

proposed change will 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.

Reduction of the RCS specific activity limit allows an increase

in the MSLB allowable primary to secondary leakage. The net effect

is no reduction in the margin of safety provided by 10 CFR part 100

and GDC 19 criteria. The maximum allowable leakage is the leakage

limit for projected SG leakage following SG tube inspection and

repair. Reducing specific activity to increase projected leak rate

follows guidance given by GL 95-05 and effectively takes margin

available in the specific activity limits and applies it to the

projected SG leak rate. This has been determined to be an acceptable

means for accepting higher projected leak rates while still meeting

the applicable limits of 10 CFR part 100 and GDC 19 criteria with

respect to offsite and control room doses. Additionally, monitoring

of the specific activity and compliance with the required actions

remains unchanged. Therefore, the proposed change does not involve a

significant reduction in the margin of safety.

For consistency, the value of secondary coolant activity in

Table TS 4.1.2 is being corrected from 1.0 microcurie/gram to 0.1

microcurie/gram. This is consistent with a previously submitted and

approved amendment, therefore, no significant hazards exist for this

change.

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 Wisconsin, Cofrin

Library, 2420 Nicolet Drive, Green Bay, WI 54311-7001

Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,

P.O. Box 1497, Madison, WI 53701-1497

NRC Project Director: Richard P. Savio

Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee

Nuclear Power Plant, Kewaunee County, Wisconsin

Date of amendment request: April 15, 1998.

Description of amendment request: The revisions in the proposed

Technical Specification amendment are part of the licensee's fuel and

reload change plan for Cycle 23. The revisions implement changes

associated with a new fuel design and also reflect changing plant

conditions due to steam generator tube plugging and repair. The

Technical Specifications (TS) would be modified as follows:

(1) Figure 2.1-1 would be revised to reflect the recently approved

High Thermal Performance (HTP) Critical Heat Flux (CHF) correlation and

corresponding Departure from Nucleate Boiling Ratio (DNBR) limit of

1.14. The figure would also reflect changes in peak rod power and

minimum reactor coolant flow.

(2) TS 3.10.b--new hot channel factors would be incorporated for

the new fuel design and the corresponding increase in peaking factors.

The limits for Height Dependent Nuclear flux Hot Channel Factor are

specified in TS 3.10.b.1 and the limits for Nuclear Enthalpy Rise Hot

Channel Factor are specified in 3.10.b.2.

(3) TS 3.10.k--the specification for the maximum Reactor Coolant

System (RCS) Inlet Temperature would be replaced with a specification

for the maximum Reactor Coolant System (RCS) Average Temperature.

(4) TS 3.10.l--the statement ``During 100% steady-state power

operation'' would be revised in the specification for minimum Reactor

Coolant System (RCS) pressure and replaced with ``During steady-state

power operation.''

(5) TS 3.10.m--the minimum Reactor Coolant Flow is being decreased

to 85,500 gallons per minute per loop.

(6) TS 3.10.n--would be revised to reflect the new Minimum DNBR

limit.

(7) Figure TS 3.10-1--the Required Shutdown Reactivity vs. Boron

Concentration would be revised to reflect the change to an 18 month

fuel cycle.

(8) Figure TS 3.10-2, the Hot Channel Factor Normalized Operating

Envelope would be revised to reflect the values used in the new safety

analyses.

(9) The Table of Contents and the Basis sections would be revised

to accommodate the above 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:

Figure TS 2.1-1: The proposed changes will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The safety limits curves are not accident initiators. Therefore,

the change will not increase the probability of an accident

previously evaluated. The proposed changes to the safety limits

curves do not alter the plant configuration, operating set points,

or overall plant performance. The safety limits curves reflect the

changes to the DNBR limit, CHF correlation, RCS flow peaking factors

and fuel design. The significant hazards determinations for these

parameters are evaluated later in this submittal. Therefore, the

change will not increase the consequences of an accident previously

evaluated.

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

from any accident previously evaluated.

The proposed changes in the safety limits curves do not alter

the plant configuration, operating set points, or overall plant

performance. Therefore, it does not create the possibility of a new

or different kind of accident.

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

Operation in the acceptable regions (i.e., below and to the left

of the safety limit curves) in combination with the reactor

protection and engineered safety systems designed into the plant

will ensure that the safety limits are not exceeded during normal

operation or during anticipated design basis operational transients.

The core will be operated in the nucleate boiling heat transfer

regime. Departure from nucleate boiling (DNB) will not occur and

therefore fuel cladding integrity will be assured.

The revised safety limit curves have been developed using

operating parameters at their bounding values (e.g., rod powers at

the peaking factor limits, reactor coolant flow at the minimum

operating limit). The revised curves will bound plant operation with

Siemens Power Corporation standard or heavy fuel. Therefore, this

change will not involve a significant reduction in safety margin.

TS 3.10.b: The proposed changes will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

Peaking factor limits are input assumptions to the safety

analyses and are not accident initiators. Therefore, this change

would not increase the probability of occurrence of an accident

previously evaluated.

The safety analyses input assumptions are designed to bound

actual plant operation. Changing the safety analysis input

assumption for the increased peaking factor limits does not change

the underlying progression of design basis accidents evaluated in

the safety analyses. All safety analysis acceptance criteria are

satisfied in the increased peaking factor limit conditions.

Additionally, the radiological consequences

[[Page 25121]]

are bounded by existing analysis at the increased peaking factor

limits. Therefore, this change will not significantly increase the

consequences of an accident previously analyzed.

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

from any accident previously evaluated.

This change incorporates the safety analyses assumptions for

core peaking factor limits for Siemens Power Corporation heavy fuel.

The change does not alter plant equipment, set points or plant

performance. Therefore, changing the peaking factor limits for

analysis purposes will not create a new or different kind of

accident from any accident previously evaluated.

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

Results of the safety analyses and of radiological consequences

indicate that all acceptance criteria are satisfied. The peaking

factor limits assumed in the safety analyses are consistent with the

proposed revised limits and these revised limits are established to

bound actual plant operation. Therefore, this change will not

involve a significant reduction in the margin of safety.

TS 3.10.k: The proposed change will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The RCS average temperature limit is not an accident initiator.

Changing the technical specification limit consistent with the

accident analyses will not increase the probability of an accident

previously evaluated.

The proposed change limits the maximum reactor coolant system

average temperature to 568.8 deg.F. The design basis safety

analyses, the Large and Small Break LOCA accidents and the non-LOCA

accidents, have been analyzed and/or evaluated consistent with the

revised RCS average temperature. The re-analysis and evaluation have

demonstrated that all safety analysis acceptance criteria are

satisfied at the specified temperature. Therefore, the change will

not increase the consequences of an accident previously evaluated.

The proposed technical specification limit for maximum allowed

RCS average temperature was decreased below the analytical limit to

account for instrument error.

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

from any accident previously evaluated.

The proposed change does not alter the plant configuration,

operating set points, or overall plant performance. Therefore, it

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

accident.

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

The proposed change is consistent with the safety analyses. All

safety analyses acceptance criteria are satisfied at the revised

reactor coolant system average temperature. The TS limit will bound

actual plant operation. Therefore, there is no significant reduction

in the margin of safety.

TS 3.10.l: The proposed change will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The RCS pressure limit is not an accident initiator. By removing

the 100% value from the specification, the assumptions in the safety

analyses are not changed. Changing the technical specification to

remove the 100% power criteria will not increase the probability of

an accident previously evaluated.

The design basis safety analyses have been analyzed and/or

evaluated at the specified RCS pressure. The analyses and

evaluations have demonstrated that all safety analyses acceptance

criteria are satisfied at this pressure. Therefore, the change would

not increase the consequences of an accident previously evaluated.

The proposed technical specification limit for minimum allowed

RCS pressure was increased above the analytical limit to account for

instrument error.

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

from any accident previously evaluated.

The proposed change does not alter the plant configuration,

operating set points, or overall plant performance. Therefore, it

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

accident.

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

The proposed change is consistent with the safety analyses. All

safety analyses acceptance criteria are satisfied at the reactor

coolant system pressure. The limit will bound actual plant

operation. Therefore, there is no significant reduction in the

margin of safety.

TS 3.10.m: The proposed change will not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The RCS flow limit is not an accident initiator. Changing the

technical specification limit consistent with the accident analysis

will not increase the probability of an accident previously

evaluated.

The proposed change limits the minimum reactor coolant flow. The

design basis safety analyses have been analyzed and/or evaluated at

the revised RCS flow. The re-analysis and evaluation have

demonstrated that all safety analysis acceptance criteria are

satisfied at the specified flow. Therefore, the change will not

significantly increase the consequences of an accident previously

evaluated.

The proposed technical specification limit for minimum

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