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

NUCLEAR REGULATORY COMMISSION

Biweekly Notice; Applications and Amendments to Facility

Operating Licenses Involving No Significant Hazards Considerations

I. Background

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

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

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

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

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

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

the Commission the authority to issue and make immediately effective

any amendment to an operating license upon a determination by the

Commission that such amendment involves no significant hazards

consideration, notwithstanding the pendency before the Commission of a

request for a hearing from any person.

This biweekly notice includes all notices of amendments issued, or

proposed to be issued from December 7, 1998, through December 17, 1998.

The last biweekly notice was published on December 16, 1998 (63 FR

69332).

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

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different kind of accident from any accident previously evaluated; or

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

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

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

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

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

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

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

the Commission may issue the license amendment before the expiration of

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

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

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

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

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

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

Directives Branch, Division of 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 January 29, 1999, 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).

[[Page 71964]]

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

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

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

Carolina

Date of amendment request: October 27, 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 Operating Licenses for the

BSEP units. The amendments are administrative in nature and would

delete various completed license conditions, make editorial changes,

and provide clarifying information.

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.

Basis for a 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 changes revise the BSEP, Unit Nos. 1 and 2, Facility

Operating Licenses to delete various license conditions that have been

completed, make editorial changes, and provide clarifying information.

The changes are administrative and only provide updated and clarifying

information. No physical or operational changes to the facility will

result from the proposed changes. 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.

The proposed changes revise the BSEP, Unit Nos. 1 and 2, Facility

Operating Licenses to delete various license conditions that have been

completed, make editorial changes, and provide clarifying information.

The changes are administrative and only provide updated and clarifying

information. The proposed license amendments do not alter any plant

operation and will not result in a physical change to the facility.

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 changes revise the BSEP, Unit Nos. 1 and 2, Facility

Operating Licenses to delete various license conditions that have been

completed, make editorial changes, and provide clarifying information.

The changes are administrative and only provide updated and clarifying

information. No physical or operational changes to the facility will

result from the proposed changes. 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

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: Frederick J. Hebdon.

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

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

Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, Units 1 and

2, Rock Island County, Illinois

Date of application for amendment request: November 30, 1998.

Description of amendment request: This amendment request proposes

to relocate, to a licensee controlled document, the requirement for

removal of the Reactor Protection System (RPS) shorting links. Removal

of the shorting links enables a non-coincident scram on high neutron

flux as detected by the Source Range Monitors (SRMs).

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?

The RPS shorting links are not precursors to any previously

evaluated accident. The Source Range Monitors (SRMs), and the ability

of the SRMs to provide a RPS trip, are also not precursors to any

previously evaluated accident. Therefore, relocating the RPS shorting

link requirement to administrative controls [the Updated Final Safety

Analysis Report (UFSAR)] will not increase the probability of an

accident previously evaluated.

The RPS shorting links are not assumed to be removed in any

accident analysis, and the SRMs are not assumed to provide a RPS trip

in any accident analysis. The refueling interlocks and SHUTDOWN MARGIN

calculations will continue to provide assurance of reactivity control.

Therefore, relocating the RPS shorting link requirements to

administrative controls [the UFSAR] will not increase the consequences

of an accident previously evaluated.

The RPS shorting link requirements will be relocated to

administrative controls that are administered pursuant to the

requirements of 10 CFR 50.59, thereby reducing the level of regulatory

control. The level of regulatory control has no impact on the

probability or consequences of an accident previously evaluated.

Consequently, 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?

Relocating the RPS shorting link requirements to administrative

controls [the UFSAR] does not create any new failure mechanisms. No new

equipment will be installed or utilized, and no new operating

conditions will be initiated as a result of this change. Therefore, the

proposed change does 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?

The refuel interlocks and SHUTDOWN MARGIN calculations will

continue to ensure that the reactor stays

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subcritical in the Refuel Mode. The margin to safety as represented by

the SHUTDOWN MARGIN designed into the core and verified in the SHUTDOWN

MARGIN calculations will be unaffected by relocation of the RPS

shorting link requirements to administrative controls [the UFSAR]. The

margin to safety as represented by the fuel bundle drop assumptions

protected by the refuel interlocks will be unaffected. In addition, no

accident analysis assumes that the RPS shorting links are removed. In

addition, the RPS shorting link requirements will be relocated to

administrative controls [the UFSAR] for which future change will be

evaluated pursuant to the requirements of 10 CFR 50.59. Therefore,

there will be no change in the types or significant increase in the

amounts of any effluents released offsite, and, thus, 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

amendments requested involve no significant hazards consideration.

Local Public Document Room location: for Dresden, Morris Area

Public Library District, 604 Liberty Street, Morris, Illinois 60450;

for Quad Cities, 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.

Florida Power Corporation, et al. (FPC), Docket No. 50-302, Crystal

River Nuclear Generating Plant, Unit No. 3 (CR-3), Citrus County,

Florida

Date of amendment request: October 30, 1998 (LAR-236).

Description of amendment request: The proposed amendment would

change the Crystal River Unit 3 (CR-3) Improved Technical

Specifications (ITS) Section 5.6.2.19, Section 3.4.11, Bases 3.4.11 and

Bases 3.4.3. The changes reflect the use of fluence methodology

described in Topical Report BAW-2241P, ``Fluence and Uncertainty

Methodologies,'' and the use of American Society of Mechanical

Engineers (ASME) Code Case N-514, ``Low Temperature Overpressure

Protection,'' for developing Low Temperature Overpressure Protection

(LTOP) limits. Reference to Topical Report BAW-1543A, ``Integrated

Reactor Vessel Surveillance Program,'' was also added to ITS Section

5.6.2.19. ITS Section 3.4.11 (Low Temperature Overpressure Protection

System), was revised to reflect the new LTOP limits based on revised

fluence projections through 32 Effective Full Power Years (EFPY). The

Pressure/Temperature (P/T) Limits Report is being revised to reflect

the new P/T limits for heatup, cooldown, hydrostatic and leak test, and

to incorporate the CR-3 LTOP curve.

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.

LAR [License Amendment Request] #236 proposes several changes to

the ITS operational limits. These changes are being proposed to

maintain the necessary margins of safety through 32 EFPY using analyses

based on methodologies that have been previously approved for use at

CR-3, ASME Code Case N-514 and LTOP SER [Safety Evaluation Report], and

are currently being reviewed by the NRC staff:

--NRC to FPC letter, 3N1293-30, dated December 20, 1993, ``Crystal

River Unit 3--Issuance of Amendment RE: Improved Technical

Specifications (TAC No. M74563)''

--NRC to FPC letter, 3N1297-16, dated December 22, 1997, ``Crystal

River Unit 3--Staff Evaluation and Issuance of Amendment RE: Low-

Temperature Overpressure Protection (TAC No. M99277)''

--NRC to FPC letter, 3N079705, dated July 3, 1997, ``Crystal River 3--

Exemption from Requirements of 10 CFR 50.60, Acceptance Criteria for

Fracture Prevention for Lightwater Nuclear Power Reactors for Normal

Operation (TAC No. M98380)''

--BAW-2241P, ``Fluence and Uncertainty Methodologies''

The limiting transient for LTOP remains a failed-open makeup valve.

Existing LTOP controls (maximum of one makeup pump capable of injecting

into the RCS [reactor coolant system], high pressure injection (HPI)

deactivated, the CFTs [core flood tanks] isolated, pressure relief

capability and maintaining a gas volume in the RCS) remain unchanged

from the current ITS 3.4.11 as approved by Reference 3, except the

setpoints proposed herein. The setpoints are being updated to reflect

the new 32 EFPY fluence analysis and P/T limits. Therefore, this change

does not involve a significant increase in the probability or

consequences of any 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 create the possibility of a new or

different kind of accident from any previously evaluated since they do

not introduce new systems, failure modes or plant perturbations.

Therefore, this change does not create the possibility of a new or

different kind of accident from any previously evaluated.

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

The proposed changes will not involve a significant reduction in

the margin of safety since the proposed P/T limitations have been

developed consistent with the requirements of 10 CFR 50.60. The

operational limits have been developed to maintain the necessary

margins of safety as defined by ASME through 32 EFPY using

methodologies previously reviewed and approved by the NRC. The

objective of these limits is to prevent non-ductile failure during any

normal operating condition, including anticipated operational

occurrences and system hydrostatic tests.

The LTOP safety factors are based on reanalyzed conditions for 32

EFPY of operation utilizing methodology contained in ASME Code Case N-

514 which has been approved for use at CR-3. The Code Case provides an

acceptable margin of safety against flaw initiation and reactor vessel

failure. The application of Code Case N-514 for CR-3 ensures an

acceptable level of safety. Therefore, this 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 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 14024, St. Petersburg, Florida

33733-4042.

NRC Project Director: Frederick J. Hebdon.

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

Nuclear Generating Plant, Unit No. 3 (CR-3), Citrus County, Florida

Date of amendment request: October 30, 1998.

Description of amendment request: The proposed amendment requests

approval of a change to the Crystal River Unit 3 (CR-3) Final Safety

Analysis Report (FSAR) regarding the methodology for performing the

Spent Fuel Pool (SFP) B criticality analysis. Recent Boraflex samples

from the SFP B demonstrate a weight loss in excess of the available

margin within the current licensing basis calculation. The criticality

analysis calculations proposed in this amendment request demonstrate

that the burnup/enrichment curves in the current Improved Technical

Specifications (ITS) have sufficient margin to accommodate up to a 20%

loss in Boraflex neutron absorption, and still maintain SFP B at less

than or equal to 0.95 k-effective when fully loaded and flooded with

unborated water. Florida Power Corporation has concluded that the

change in the criticality analysis methodology represents an unreviewed

safety question, and thus requires prior NRC approval.

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.

No. The two possible accidents are: (1) criticality during normal

storage and (2) criticality due to a misloaded fuel assembly during

handling fuel. Each are discussed below:

(1) Criticality during normal storage.

For criticality during normal storage to occur, there must be a

loss of negative reactivity since an addition of positive reactivity is

not possible without fuel movement. A loss in negative reactivity could

result only from reduction in Boraflex inventory below that needed to

meet the design basis. The proposed criticality analysis for Spent Fuel

Pool B demonstrates that Spent Fuel Pool B is capable of maintaining

the design basis requirement of k-effective less than or equal to 0.95

when flooded with unborated water and with a loss of up to 20% of the

Boraflex absorber material. Therefore, allowing up to 20% Boraflex loss

with the new analysis does not significantly increase the probability

of an accident previously evaluated.

(2) Criticality during fuel handling.

Criticality during fuel handling could occur due to loss of

negative reactivity, or the addition of positive reactivity. Loss of

negative reactivity could result from loss of Boraflex as discussed

above.

Addition of positive reactivity would result from the misloading of

fuel in a fashion not in accordance with ITS LCO 3.7.15, such as the

misloading of a fresh 5.05% enriched fuel assembly into Region 2 or

side-by-side with another fresh fuel assembly in Region 1. The minimum

required boron concentration of ITS LCO 3.7.14 and CR-3 FSAR 9.3.2.1.2

are intended to compensate for just such an accident. Consistent with

the double-contingency principle, a boron dilution is not required to

be considered concurrent with a misloaded new fuel assembly (bases of

ITS LCO 3.7.14). The use of a new calculational method will not

increase the probability of fuel assembly misloading. A boron dilution

event without an accompanying misloaded fuel assembly is not impacted

by the new criticality analysis, since the design basis allows for

unborated water for normal storage conditions.

Therefore, since the proposed criticality analysis does not

increase the probability of a misloaded fuel assembly, the probability

of an occurrence of an accident previously evaluated is not

significantly increased.

Boraflex is credited with preventing inadvertent criticality. It is

not credited with mitigating the effects, or dose consequences, to the

public or to plant personnel from an inadvertent criticality. The

criticality analysis does not affect or mitigate the dose consequences

to the public or plant personnel from an inadvertent criticality.

There are no other SAR accidents that could be affected. Therefore,

the use of the proposed criticality analysis, does not significantly

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.

No. The only purpose, or function, of Boraflex is reactivity

control. Therefore, the use of the proposed criticality analysis can

only result in reactivity related accidents, such as an inadvertent

criticality. Though a spent fuel pool criticality accident is not

discussed in detail, a calculation to ensure such an accident could not

occur is referenced by both FSAR 9.3 and 9.6. Therefore, this is an

accident already discussed by the SAR and dependence on a new

criticality analysis does not create the possibility of an accident of

a new or different kind than any previously evaluated.

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

No. The proposed analysis demonstrates that the safety function and

design basis are met even for a Boraflex loss of up to 20%. Though the

proposed criticality analysis methodology is more realistic, and has

been licensed at other sites, it is less conservative than the

existing, NRC approved analysis that is currently part of the CR-3

licensing basis. Additionally, it permits operation with a greater loss

of Boraflex than the existing analysis.

The current licensing basis, BAW-2209, ``Crystal River Unit 3 Spent

Fuel Storage Pool Criticality Analysis'', provides the analytical basis

of both ITS LCO 3.7.14 and LCO 3.7.15. This analysis uses very

conservative assumptions and methodologies, and results in very little

margin remaining for identified Boraflex loss. The margin of safety,

although less than previously evaluated, is not significantly reduced

with reliance on the current criticality analysis. The margin of safety

is restored with use of the proposed criticality analysis. Therefore,

the margin of safety is not significantly reduced with use of the

proposed criticality analysis.

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: 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 (CR-3), Citrus County, Florida

Date of amendment request: November 23, 1998.

Description of amendment request: The proposed amendment would

change the CR-3 Improved Technical Specifications (ITS) to raise the

Engineered Safeguards Actuation System (ESAS) setpoint for reactor

coolant system (RCS) low pressure from

[[Page 71967]]

1500 psig to 1625 psig. This change is intended to provide for earlier

actuation of high pressure injection (HPI) following certain small

break loss of coolant accidents and result in a lower peak center line

temperature (PCT) during these transients. The applicability

requirement for ESAS operability would be changed from greater than

1700 psig to greater than 1800 psig to maintain the previous margin

above the ESAS setpoint. Similarly, the reactor protection system (RPS)

setpoint for RCS low pressure and the RPS setpoint for Shutdown Bypass

(RCS High Pressure) would each be raised by 100 psig to maintain the

previous pressure margins. In addition, Surveillance Requirement

3.5.2.5 would be revised such that valves in the HPI flowpath that are

throttled to balance flow between the four HPI lines would be verified

in the correct position. The need for these changes resulted from

planned modifications to the HPI system to improve performance and

reliability of this system. Changes to ITS Bases necessitated by the

system modifications and setpoint changes are included in the

submittal.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below.

1. Does not involve a significant increase in the probability or

consequences of an accident previously evaluated.

The setpoint changes for reactor trip and High Pressure Injection

(HPI) actuation will result in a very small (approximately one-percent)

increase in the probability for reactor trips. Review of industry data

shows that this increase is not significant. The revised accident

analysis has determined that transients which reduce Reactor Coolant

System (RCS) pressure below the new setpoints, warrant the associated

action. Engineered Safeguards Actuation System (ESAS) and Reactor

Protection System (RPS) actuations are used to mitigate accidents and

are not the initiator of analyzed accidents. Therefore, the probability

of previously evaluated accidents is not affected.

RPS and ESAS functions are assumed to actuate to mitigate

transients. The revised setpoints will ensure earlier actuation of the

RPS and ESAS on a low RCS pressure condition. Raising the ESAS Low RCS

Pressure Setpoint will ensure earlier automatic HPI actuation for a

portion of the spectrum of pressure decreasing events. For rapid

depressurization events, such as main steam line break and large break

Loss of Coolant Accident (LOCA), this will have little impact. For

slower events, or those that do not reach the current setpoint during

the initial subcooled blowdown phase, HPI will be automatically

initiated substantially earlier in the event. This will increase the

integrated HPI flow to the RCS during the time the core is likely to be

uncovered, thereby reducing the consequential PCT. This additional flow

results in a significant peak clad temperature (PCT) decrease for small

break LOCA scenarios less than 0.07 square feet. Based on the above,

the consequences of previously evaluated accidents will not be

increased.

The HPI system characteristics will not be affected such that the

probability of any accident is increased. The system flow restriction

for protection from low temperature overpressure (LTOP) events will be

maintained. The HPI system is used for accident mitigation and is not

the initiator of evaluated accidents other than LTOP. The proposed

surveillance changes will ensure that all valves throttled in the HPI

flowpath are verified and secured in the correct position. The throttle

valves and stop check valves will be positioned to ensure HPI flow is

within analyzed limits. Therefore, the consequences of accidents that

rely on HPI flow will not be increased.

Based on the above evaluation, the probability or consequences of

evaluated accidents are not significantly increased by these changes.

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

accident from any accident previously evaluated.

The change to RPS and ESAS setpoints will not change the functions

of plant equipment, no new system interactions will be created, and no

new failure modes will be introduced. The setpoint changes will permit

earlier actuation for the associated actions. However, no new plant

conditions will be introduced by the setpoint changes.

The HPI modifications include the installation of throttle valves

that will change the flow characteristics of the system. The new

throttle valves are manual valves that will be secured in position. The

revised surveillance requirements will ensure these valves are

positioned such that HPI flow is within analyzed limits. Therefore, no

conditions are created that could cause a new type of accident.

Based on the above evaluation, these changes cannot create the

possibility of an accident of a different type than previously

evaluated in the [Safety Analysis Report] SAR.

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

safety.

The safety function of the affected portions of the RPS and ESAS

systems is to actuate their respective functions if RCS pressure drops

below the setpoint. The raised RPS and ESAS setpoints will provide

earlier actuation for these protective features. These changes will

increase the margin of safety provided by the associated Technical

Specifications.

The safety function of the HPI system is to provide cooling to

limit fuel peak clad temperature. The revised surveillance requirements

will ensure valves are positioned such that HPI flow is within analyzed

limits. Therefore, the margin of safety provided by the HPI

surveillance requirements is maintained.

Based on the above evaluation, there is no reduction in the margin

of safety associated with the equipment and systems affected by this

change.

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: 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.

GPU Nuclear, Inc., et al., Docket No. 50-289, Three Mile Island Nuclear

Station, Unit No. 1, Dauphin County, Pennsylvania

Date of amendment request: December 3, 1998.

Description of amendment request: The proposed change revises the

TMI-1 Core Protection Safety Limits and Core Protection Safety Bases,

as specified in Technical Specification Figures 2.1-1 and 2.1-3, to

provide more restrictive limits which reflect the decrease in reactor

coolant system flow resulting from the analysis of increased once-

through steam generator (OTSG) tube plugging limits (total allowable

number of tubes plugged). The licensee is currently restricted to a

total of 2,000 tubes plugged in both OTSGs which corresponds to 6.4

percent of the total number of tubes. The licensee's more restrictive

Core Protection Safety Limits reflect the reduction in reactor coolant

[[Page 71968]]

flow that would exist if an average of 20 percent of the OTSG tubes

were plugged.

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. Operation of the facility in accordance with the proposed

amendment would not involve a significant increase in the probability

of occurrence or the consequences of an accident previously evaluated.

An increase in the average steam generator tube plugging (SGTP) level

to 20% results in a small reduction of reactor coolant system (RCS)

flow rates and primary to secondary heat transfer. These changes result

in small changes to the primary and secondary side operating

parameters, and do not result in any additional challenges to plant

equipment. The proposed Technical Specification Changes resulting from

the increase in allowable tube plugging limits are more restrictive but

remain bounded by the existing reactor protection system (RPS) trip

setpoints. The assessment of the NSSS [nuclear steam supply system]

primary components, including the reactor pressure vessel, reactor

core, reactor coolant pump, steam generator, pressurizer, control rod

drive mechanisms, and RCS piping concluded that the integrity of these

components will be unaffected by the increase in average SGTP level.

A re-analysis of the bounding Updated Final Safety Analysis Report

(UFSAR) Chapter 14 accidents, specifically the startup accident, loss

of coolant flow, loss of feedwater, and large and small break LOCA

demonstrated compliance with the acceptance criteria. The RCS pressure

boundary is not challenged, and the DNBR [departure from nucleate

boiling ratio] and peak clad temperature values remain within the

specified limits of the licensing basis. An analysis of the loss of

electric power accident demonstrated the ability of the plant to

transition smoothly to natural circulation with an average of 20% SGTP

or with asymmetric plugging. It was also determined that the current

mass and energy release data used for the containment integrity and

equipment qualification remain bounding. Since the design requirements

and safety limits continue to be met, system functions are not

adversely impacted, and the integrity of the RCS pressure boundary is

not challenged, the radiological consequences remain unchanged.

Therefore, this activity does not involve a significant increase in the

probability of occurrence or the consequences of an accident previously

evaluated.

2. Operation of the facility in accordance with the proposed

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

of accident from any previously evaluated. The proposed Technical

Specification changes are more restrictive core protection safety

limits but remain bounded by the existing RPS trip setpoints. This

proposed change assures safe operation commensurate with the effects of

steam generator tube plugging. This increase in the average level of

SGTP to 20% will not introduce any new accident initiator mechanisms.

No new failure modes or limiting single failures have been identified.

Since the safety and design requirements continue to be met and the

integrity of the RCS pressure boundary is not challenged, no new

accident scenarios have been created. This change does not add any new

equipment, modify any interfaces with existing equipment, or change the

equipment function or the method of operating the equipment. Reactor

core, RCS, and steam generator parameters remain within appropriate

design limits during normal operation. Therefore, this activity does

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

any previously evaluated.

3. Operation of the facility in accordance with the proposed

amendment would not involve a significant reduction in a margin of

safety. The existing RPS trip setpoints bound the proposed Technical

Specification changes resulting from 20% SGTP. This change assures safe

operation commensurate with the effects of steam generator tube

plugging. The TMI-1 DNB design basis, RCS pressure limits, peak clad

temperature limits and dose criteria are maintained for all UFSAR

transients. Therefore, this activity does not reduce the margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Law/Government Publications

Section, State Library of Pennsylvania (REGIONAL DEPOSITORY), Walnut

Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.

Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,

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

NRC Project Director: Cecil O. Thomas.

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point

Nuclear Station, Unit 2 (NMP2), Oswego County, New York

Date of amendment request: November 16, 1998.

Description of amendment request: The proposed amendment would

revise Technical Specifications (TSs) related to the implementation of

systems for the detection and suppression of coupled neutronic/thermal-

hydraulic instabilities in the reactor. Average Power Range Monitor

(APRM) flow control trip reference cards will initiate a reactor scram

to limit the oscillation magnitude at reactor trip so as to limit the

associated Critical Power Ratio change and, in conjunction with Minimum

Critical Power Ratio (MCPR) operating limits, assure compliance with

the MCPR safety limit. In addition, the changes would increase the APRM

flow biased neutron flux scram and control rod block settings to allow

plant operation in the Extended Load Line Limit Analysis region. Thus,

the proposed changes are in regard to setpoints and calculations for

fuel cladding integrity and the associated TS Bases. In the Bases for

TS 2.1.1, the proposed change would reference new equations in TS

2.1.2a. In TS 2.1.2a, the proposed change would be to the equation for

determining the flow biased APRM scram and rod block trip setpoints. In

the Bases for TS 2.1.2a, the proposed change would reflect the new

setpoints. In the Bases for TS 2.2.2, the proposed change would be to

the description of the setpoint methodology which is based upon General

Electric Report NEDC-31336, ``GE Instrumentation Setpoint

Methodology.'' In Note (m) of TS Table 3.6.2/4.6.2, the proposed change

would be to the calibration range for the APRM channel setpoint. In the

Bases for TS 3.6.2/4.6.2, the proposed change would be to the equations

and methodology for determining APRM scram and rod block setpoints. In

TS 6.9.1.f, which identifies documents approved by NRC for analytical

methods used to determine core operating limits, the proposed change

would add ``NEDO-32465-A, Reactor Stability Detect and Suppress

Solutions Licensing Basis Methodology for Reload Applications, August

1996.''

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

[[Page 71969]]

consideration, which is presented below:

The operation of Nine Mile Point Unit 1, in accordance with the

proposed amendment, will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The APRM neutron monitoring system is not an initiator or a

precursor to an accident. The neutron monitoring system monitors the

power level of the reactor core and provides automatic core protection

signals in the event of a power transient. A Restricted Region will be

maintained such that the probability of a stability event is not

increased. Therefore, the proposed TS changes cannot affect the

probability of a previously evaluated accident.

The proposed TS changes will revise the APRM flow-biased neutron

flux scram TS setting to provide automatic protection to assure that

anticipated coupled neutronic/thermal-hydraulic instabilities will not

compromise established fuel safety limits. The proposed changes will

result in a more restrictive APRM flow-biased scram trip setting in the

low flow regions of the power/flow operating map (i.e., operational

conditions where reactor instabilities are most probable). In other

words, the new settings will provide a scram sooner (at a lower power

level) than the existing settings. The associated control rod block

setting will also be revised. A margin between the control rod block

and flux scram has been determined by calculation.

The proposed changes will also revise the APRM flow-biased neutron

flux scram and control rod block TS settings to provide an increase

above the current values in operating conditions not susceptible to

reactor instabilities. Specifically, the proposed changes will

implement a 2% increase in the analytical limit of the APRM flow-biased

flux scram and a 7% increase in the analytical limit of the APRM flow-

biased control rod block. Evaluation demonstrates that these proposed

analytical limit increases have negligible impact on the transient

events results for NMP1 [Nine Mile Point Unit 1] as documented in

Chapter XV of the NMP1 UFSAR, [Updated Final Safety Analysis Report],

including the limiting transient events which are reanalyzed each

reload. Of the twenty-five (25) transient events analyzed in Section XV

of the NMP1 UFSAR, only the Inadvertent Startup of Cold Recirculation

Loop event and the Recirculation Flow Controller Malfunction--Increase

Flow event have potentially impacted results. The Chapter XV Control

Rod Drop Accident as well as the Turbine Trip with No Bypass at Partial

Power event were also evaluated.

For the Inadvertent Startup of Cold Recirculation Loop event, the

proposed 2% increase in the high neutron flux scram would result in an

increase in the fuel average surface heat flux response. However, there

is significant margin between the surface heat flux value for this

event and the current limiting MCPR [Minimum Critical Power Ratio]

event (the Feedwater Controller Failure Maximum Demand event). As such,

any small change to the fuel surface heat flux response due to the high

neutron flux scram analytical limit increase would not result in the

fuel thermal margin requirements for the Inadvertent Startup of Cold

Recirculation Loop event to exceed the MCPR limits set by the limiting

reload analysis event.

The reactor neutron flux for the Recirculation Flow Controller

Malfunction--Increase Flow event also showed an increasing trend from

its initial value. However, the peak response for this parameter (104%

of rated) is significantly below the high neutron flux scram analytical

limit. Accordingly, the proposed increase to the high neutron flux

scram analytical limit does not affect the response to this transient

event.

The Control Rod Drop Accident is included in Chapter XV of the NMP1

UFSAR. As noted in NEDE-24011-P-A, ``GESTAR II: General Electric

Standard Application for Reactor Fuel,'' the initial power burst from

this event is terminated by the Doppler reactivity feedback while the

scram provides the final event termination several seconds later. The

120% APRM scram limit was conservatively chosen. The time delay

introduced by the small change in analytical limit will be

inconsequential due to the extremely rapid power rise for this event

(i.e., the time of scram for a 120% analytical limit vs. a 122%

analytical limit is essentially the same).

The proposed Bases changes to TS 3.6.2/4.6.2 and TS 2.2.2 simply

provide details of the setpoint methodology currently used as well as

specific allowable values.

Therefore, the proposed TS changes to implement a more restrictive

flow-biased scram setting to protect against reactor instabilities and

the proposed change to increase the high neutron flux scram and rod

block analytical limits do not result in a significant increase in the

consequences of an accident previously evaluated.

The operation of Nine Mile Point Unit 1, in accordance with the

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

different kind of accident from any accident previously evaluated.

The proposed changes will revise the APRM flow-biased neutron flux

scram TS settings to assure anticipated coupled neutronic/thermal-

hydraulic instabilities will not compromise established fuel safety

limits in the low flow regions of the power/flow operating map as well

as revise the associated control rod block settings. These changes also

propose a 2% increase in the analytical limit of the APRM flow-biased

neutron flux scram and a 7% increase in the analytical limit of the

APRM flow-biased control rod block. These changes do not introduce any

new accident precursors and do not involve any alterations to plant

configurations which could initiate a new or different kind of

accident. The proposed changes do not affect the intended function of

the APRM system nor do they affect the operation of the system in a way

which would create a new or different kind of accident.

Therefore, the proposed changes will not create the possibility of

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

The operation of Nine Mile Point Unit 1, in accordance with the

proposed amendment, will not involve a significant reduction in a

margin of safety.

More conservative APRM flow-biased neutron flux scram and control

rod block settings will be implemented in the low flow regions of the

power/flow operating map. The scram setting change will assure that

anticipated coupled neutronic/thermal-hydraulic instabilities will not

compromise established fuel safety limits. The proposed changes will

also implement a 2% increase in the APRM flow-biased neutron flux scram

and a 7% increase in the APRM flow-biased control rod block in those

operating regions not susceptible to reactor instabilities. Evaluation

demonstrates that these proposed increases have negligible impact on

the transient events or accident results for NMP1. The impacted

transient events are either not the limiting MCPR event, the peak

response to the event is significantly below the high neutron flux

scram analytical limit or in the case of the Control Rod Drop Accident,

the time delay introduced by the change will be inconsequential due to

the extremely rapid power rise. No other events are adversely affected.

Therefore, the proposed amendment does not involve a significant

reduction in a margin of safety.

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

this

[[Page 71970]]

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

Therefore, the NRC staff proposes to determine that the amendment

request involves no significant hazards consideration.

Local Public Document Room location: Reference and Documents

Department, Penfield Library, State University of New York, Oswego, New

York 13126.

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

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

NRC Project Director: S. Singh Bajwa.

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point

Nuclear Station, Unit 2 (NMP2), Oswego County, New York

Date of amendment request: November 19, 1998.

Description of amendment request: The proposed amendment would

change the surveillance frequencies in Technical Specifications (TSs)

4.8.4.4a, ``Surveillance Requirements--Reactor Protection System

Electric Power Monitoring (RPS Logic),'' and 4.8.4.5a, ``Surveillance

Requirements--Reactor Protection System Electric Power Monitoring

(Scram Solenoids),'' to require channel functional testing of the RPS

Motor Generator Set (M/G) and RPS Uninterruptible Power Supplies (UPS)

Electrical Protection Assemblies (EPAs) at least once every 6 months.

These TSs currently require that channel functional testing be

performed each time the plant is in cold shutdown for a period of more

than 24 hours, unless performed within the previous 6 months.

Basis for proposed no significant hazards consideration

determination: During the last refueling outage, the licensee modified

the Nine Mile Point Unit No. 2 (NMP2) design for the RPS M/G and RPS

UPS EPAs to provide relay actuated protection systems. The relays of

the new design may be individually isolated from an essential power

circuit for testing and may be actuated without tripping the associated

breaker. The relay actuated system will allow the EPA system monitoring

an essential power supply to be functionally tested with the plant on-

line. The EPA relay actuation setpoints are not affected by the

modification or the proposed TS changes. The licensee states that the

design, installation, and testing of the new units meet the criteria of

the same standards that were applied to the previous units.

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 operation of Nine Mile Point Unit 2, in accordance with the

proposed amendment, will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The proposed changes affect surveillance testing frequency only.

The new relay actuated protection system design functions in the same

fail safe manner as the old units. Also, the new design in conjunction

with the testing capability has increased EPA reliability, while

introducing little risk to testing the EPAs with the plant in

operation. Therefore, the proposed changes to the NMP2 TS do not

involve a significant increase in the probability or consequences of an

accident previously evaluated.

The operation of Nine Mile Point Unit 2, in accordance with the

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

different kind of accident from any accident previously evaluated.

The proposed changes affect surveillance testing frequency of relay

actuated protection circuits only. The proposed changes do not

introduce any new or different accident initiators from any that were

previously evaluated. EPA relay actuation setpoints are not affected.

The actual fail safe system conditions required for EPA actuation will

remain the same. Therefore, the operation of NMP2, in accordance with

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

different kind of accident from any accident previously evaluated.

The operation of Nine Mile Point Unit 2, in accordance with the

proposed amendment, will not involve a significant reduction in a

margin of safety.

The function of the EPA systems is to isolate the loads from supply

power. That function was not altered by the proposed change.

Reliability of the EPA systems is improved. Therefore, the operation of

NMP2, in accordance with 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 50.92(c) are

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Reference and Documents

Department, Penfield Library, State University of New York, Oswego, New

York 13126.

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

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

NRC Project Director: S. Singh Bajwa.

Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile Point

Nuclear Station, Unit 1 (NMP1), Oswego County, New York

Date of amendment request: November 30, 1998.

Description of amendment request: The proposed amendment would

correct Technical Specification (TS) 3.1.2, ``Liquid Poison System,''

and the associated TS Bases. Specifically, in the Bases for TS 3.1.2,

the boron-10 concentration of 120 ppm (which is incorrectly calculated

using atomic percent instead of weight percent) would be changed to

109.8 ppm. In TS 3.1.2, the minimum volume of the sodium pentaborate

solution contained in the Liquid Poison System storage tank would be

increased from 1185 gallons to 1325 gallons.

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 operation of Nine Mile Point Unit 1, in accordance with the

proposed amendment, will not involve a significant increase in the

probability or consequences of an accident previously evaluated.

The Liquid Poison System is designed to provide the capability to

bring the reactor from a full design rating to a shutdown condition

assuming none of the control rods can be inserted. The system is

manually initiated in response to a failure of the Control Rod Drive

System to shutdown the reactor. The proposed changes revise the

required liquid poison solution volume and concentration. The proposed

changes to the Technical Specifications and the Bases require no

changes to the physical facility which could adversely affect any

accident precursors. Therefore, the proposed changes cannot

significantly increase the probability of an accident.

The proposed changes will assure that the Liquid Poison System

continues to provide the capability to shutdown the reactor during an

ATWS [Anticipated Transient Without Scram] event. In addition, the

system will continue to be capable of bringing the reactor to cold

shutdown, 3 percent delta k subcritical (0.97 keff), from a

full design rating of

[[Page 71971]]

1850 megawatts thermal assuming none of the control rods can be

inserted, and considering the combined effects of coolant voids,

temperature change, fuel doppler, and xenon and samarium. Therefore,

the change to the Technical Specifications does not significantly

increase the consequences of a previously evaluated accident.

2. The operation of Nine Mile Point Unit 1, in accordance with the

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

different kind of accident from any accident previously evaluated.

Injection of the sodium pentaborate solution into the reactor

vessel has been considered in the plant design. The proposed changes

revise the required liquid poison solution volume and concentration.

The proposed changes make no physical modification to the plant which

could create the possibility of a new or different kind of accident.

The proposed changes will maintain the capability of the Liquid Poison

System to shutdown the reactor from its full design rating assuming

none of the control rods are inserted, and considering the combined

effects of coolant voids, temperature change, fuel doppler, and xenon

and samarium. Consequently, these changes do not create the possibility

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

evaluated.

3. The operation of Nine Mile Point Unit 1, in accordance with the

proposed amendment, will not involve a significant reduction in a

margin of safety.

The proposed changes revise the required liquid poison solution

volume and concentration. The proposed changes make no physical

modification to the plant which could reduce the margin of safety.

These changes will assure compliance with the requirements of

10CFR50.62, ``Requirements for Reduction of Risk from Anticipated

Transients without Scram (ATWS) Events for Light-Water-Cooled Nuclear

Power Plants.'' In addition, these changes will maintain the capability

of the Liquid Poison System to bring the reactor from a full design

rating of 1850 megawatts thermal to greater than 3 percent delta k

subcritical (0.97 keff) assuming none of the control rods

can be inserted, and considering the combined effects of coolant voids,

temperature change, fuel doppler, xenon and samarium.

The required volume of boron-10 solution in the Liquid Poison

System storage tank includes an additional 25 percent margin beyond the

amount needed to shutdown the reactor to allow for any unexpected non-

uniform mixing. Also, the total storage tank volume of sodium

pentaborate solution incorporates 197 gallons of solution which is

unavailable for injection into the reactor vessel and a 25 gallon

margin for conservatism. Additionally, using one 30 gpm Liquid Poison

System pump, the injection time is greater than 17 minutes thereby

assuring adequate mixing. The proposed changes to the liquid poison

concentration and volume ensure the NMP1 [Nine Mile Point Unit 1]

Liquid Poison System is able to meet its safety function requirements.

Therefore, this change will not involve a significant reduction in a

margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Reference and Documents

Department, Penfield Library, State University of New York, Oswego, New

York 13126.

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

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

NRC Project Director: S. Singh Bajwa.

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

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

Connecticut

Date of amendment request: December 4, 1998.

Description of amendment request: The proposed amendment would

eliminate the need to cycle the plant and its components through a

shutdown-startup cycle by allowing the next snubber surveillance

interval to be deferred until the end of refueling outage 6 or

September 10, 1999, whichever date is earlier.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

NNECO has reviewed the proposed revision in accordance with 10 CFR

50.92 and has concluded that the revision does not involve a

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

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

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

not:

1. Involve a significant increase in the probability or consequence

of an accident previously evaluated.

The proposed change is for a one time extension to the surveillance

interval of snubber inspections required by Technical Specification

4.7.10.e. The change involves revising the calendar time for snubber

interval inspections to 36 months to coincide with the time frame of

the current cycle 6 operation.

Snubber testing experience at Millstone Unit No. 3 has shown that

historical failure rates of snubbers are low. During the third

refueling outage, after an operating cycle of approximately 22 months,

the functional testing program identified multiple Type A failures

attributed primarily to original plant construction, and resulted in a

full inspection of all Type A snubbers. The snubber inspection interval

was extended to approximately 30 months by a one-time extension to the

Technical Specifications for the fourth refueling outage and only one

Type A snubber failure was identified. Subsequent outages with

operating durations of 18 and 17 months also identified only a single

Type B failure in each outage. The results of piping stress analysis

which have been performed to assess the impact of snubbers which have

failed to meet functional test acceptance criteria have shown that

neither piping system functionality or structural integrity have ever

been compromised.

During the recent cycle 6 operation Millstone 3 has experienced an

extended midcycle shutdown, where temperature, vibration effects and

normal wear on snubbers have been minimized as compared to a normal

operating cycle. The last snubber surveillance interval inspections

were completed during this midcycle shutdown. Although the calendar

surveillance interval is impacted by this change the primary conditions

that present challenges to snubbers have not been prevalent during the

extended shutdown. Given the low failure rates of snubbers over the

last 3 surveillance intervals, and the fact the operating time of the

remainder of cycle 6 will be approximately 1 year, snubber failures are

expected to be similar to previous intervals.

Accordingly the possibility of a snubber failure leading to a

Decrease in Reactor Coolant Inventory or a Decrease in Heat Removal by

the Secondary System is not increased and there is no affect on the

probability of previously evaluated accidents.

This change does not include any physical changes to the plant and

does not affect acceptance criteria or the

[[Page 71972]]

required actions for functional failures of snubbers. Accordingly there

is no increase in the consequences of previously evaluated accidents

resulting in a Decrease in Reactor Coolant Inventory or a Decrease in

Heat Removal by the Secondary System.

Thus it is concluded that the proposed revision does not involve a

significant increase in the probability or consequence of an accident

previously evaluated.

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

from any accident previously evaluated.

This proposed revision to the surveillance interval does not change

the operation of any plant system or component during normal or

accident conditions. The proposed change extends the surveillance

interval of snubber inspections required by Technical Specification

4.7.10.e. The change involves revising the calendar time for snubber

interval inspections to coincide with the time frame of current cycle 6

operation. This change does not include any physical changes to the

plant and does not affect acceptance criteria or the required actions

for functional failures of snubbers.

Thus, this proposed revision does not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

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

The proposed change extends the surveillance interval of snubber

inspections required by Technical Specification 4.7.10.e. The change

involves revising the calendar time for snubber interval inspections to

coincide with the time frame of current cycle 6 operation. This change

does not include any physical changes to the plant and does not affect

acceptance criteria or the required actions for functional failures of

snubbers. The service life of the snubbers or parts as required by

Technical Specification 4.7.10.i will not be impacted by this change

since the required replacements have already occurred and no additional

service life dates will expire prior to September 10, 1999.

Thus, it is concluded that the proposed revision does not involve a

significant reduction in a margin of safety.

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

that the proposed revision does not involve an SHC.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center,

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

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

49 Rope Ferry Road, Waterford, Connecticut.

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

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

Connecticut.

NRC Project Director: William M. Dean.

Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. Ginna

Nuclear Power Plant, Wayne County, New York

Date of amendment request: November 24, 1998.

Description of amendment request: The proposed amendment would

revise the Ginna Station Improved Technical Specifications description

of the fuel cladding material (TS 4.2.1) and to update the list of

references provided in Specification 5.6.5 for the Core Operating

Limits Report.

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:

Evaluation of Administrative Changes

The administrative changes [related to the update of references

provided in Specification 5.6.5 for the Core Operating Limits report]

do not involve a significant hazards consideration as discussed below:

1. Operation of Ginna Station in accordance with the proposed

changes does not involve a significant increase in the probability or

consequences of an accident previously evaluated. The proposed changes

revise Administrative Controls Section 5.6.5.b to update the references

to NRC approved documents which support the analysis for the Heat Flux

Hot Channel Factor in the Core Operating Limits Report and to provide

clarification to the currently applicable methodology. It revises the

Design Features Section 4.2.1 to provide clarification of the types of

zirconium alloy filler rod material that have received previous NRC

approval and to clarify that the application shall be NRC approved.

Section 4.2.1 is revised to clarify that the analyses performed to

verify compliance with the fuel safety design bases shall be cycle

specific. As such, these changes are administrative in nature and do

not impact initiators or analyzed events or assumed mitigation of

accident or transient events. Therefore, these changes do not involve a

significant increase in the probability or consequences of an accident

previously analyzed.

2. Operation of Ginna Station in accordance with the proposed

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

accident from any accident previously evaluated. The proposed

administrative changes do not affect the manner by which the plant is

operated and no new equipment will be installed. The proposed

administrative changes will not impose any new or different

requirements. All original design and performance criteria continue to

be met, and no new failure modes have been created for any system,

component, or piece of equipment. Thus, these changes do not create the

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

previously evaluated.

3. Operation of Ginna Station in accordance with the proposed

changes does not involve a significant reduction in a margin of safety.

The proposed changes will not reduce a margin of plant safety because

the methodology has been shown to meet all applicable design criteria

and ensure that all pertinent licensing basis acceptance criteria are

met. As such, no question of safety is involved, and the changes do not

involve a significant reduction in a margin of safety.

Evaluation of Less Restrictive Changes

The less restrictive change [related to the fuel cladding material

(TS 4.2.1)] does not involve a significant hazards consideration as

discussed below:

(1) Operation of Ginna Station in accordance with the proposed

change does not involve a significant increase in the probability or

consequences of an accident previously evaluated. The Westinghouse

14 x 14 VANTAGE + fuel assemblies containing fuel rods fabricated with

ZIRLO alloy meet the same fuel assembly and fuel rod design bases as

Westinghouse 14 x 14 OFA [Optimized Fuel Assembly] fuel assemblies in

the other fuel regions. In addition, the 10 CFR 50.46 criteria will be

applied to the fuel rods fabricated with ZIRLO alloy. The use of these

fuel assemblies will not result in a change to the proposed Ginna

Westinghouse 14 x 14 OFA reload design and safety analysis limits. The

ZIRLO alloy is similar in chemical composition and has similar physical

and mechanical properties as that of Zircaloy-4. Thus the cladding

integrity is maintained and the structural integrity of the fuel

[[Page 71973]]

assembly is not affected. The ZIRLO clad fuel rods improve corrosion

resistance and dimensional stability. The use of ZIRLO does not impact

the radiological consequences of accidents previously evaluated in the

Safety Analysis. The RCS [reactor coolant system] isotopic inventory is

negligibly impacted; therefore, changes in postulated releases from the

RCS or the secondary systems are negligible. Assumptions of fuel

melting in the radiological analyses are not based on the type of fuel

cladding. For those accidents where fuel melting is postulated to occur

(control rod ejection, locked [seized] RCP rotor), the amount of fuel

undergoing melting and clad damage using ZIRLO clad is bounded by the

current values used in the Safety Analysis. Therefore, the probability

or consequences of an accident previously evaluated is not

significantly increased.

(2) Operation of Ginna Station in accordance with the proposed

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

accident from any accident previously evaluated. The Westinghouse

14 x 14 VANTAGE + fuel assemblies containing fuel rods fabricated with

ZIRLO alloy will satisfy the same design bases as that used for

Westinghouse 14 x 14 OFA fuel assemblies in the other fuel regions.

Since the original design criteria is being met, the fuel rods

fabricated with ZIRLO alloy will not be an initiator for any new

accident. All design and performance criteria will continue to be met

and no new single failure mechanisms have been created. In addition,

the use of these fuel assemblies does not involve any alterations to

plant equipment or procedures which would introduce any new or unique

operational modes or accident precursors. Therefore, the possibility

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

evaluated is not created.

(3) Operation of Ginna Station in accordance with the proposed

change does not involve a significant reduction in a margin of safety.

The Westinghouse 14 x 14 VANTAGE + fuel assemblies containing fuel rods

fabricated with ZIRLO alloy do not change the proposed Ginna

Westinghouse 14 x 14 OFA reload design and safety analysis limits. The

use of these fuel assemblies containing fuel rods fabricated with ZIRLO

alloy will take into consideration the normal core operating conditions

allowed in the Technical Specifications. For each cycle reload core,

these fuel assemblies will be specifically evaluated using approved

reload design methods and approved fuel rod design models and methods

as specified in Technical Specifications. This will include

consideration of the core physics analysis peaking factors and core

average linear heat rate effects. In addition, the 10 CFR 50.46

criteria will be applied each cycle to the fuel rods fabricated with

ZIRLO alloy. Analyses or evaluations will be performed each cycle to

confirm that 10 CFR 50.46 will be met. Therefore, the margin of safety

as defined in the Bases to the Ginna Technical Specifications is not

significantly reduced.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: Rochester Public Library, 115

South Avenue, Rochester, New York 14610.

Attorney for licensee: Nicholas S. Reynolds, Winston & Strawn, 1400

L Street, NW., Washington, DC 20005.

NRC Project Director: S. Singh Bajwa.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-

362,

San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San Diego

County, California

Date of amendment requests: November 23, 1998.

Description of amendment requests: The proposed change would revise

the Technical Specifications (TS) to (1) reinstate the log power

reactor trip at or above 4E-5% RATED THERMAL POWER (RTP); (2) reinstate

reactor trips for Reactor Coolant Flow--Low (RCS flow), the Local Power

Density--High (LPD), and the Departure from Nucleate Boiling Ratio--Low

(DNBR); (3) remove the word ``automatically'' from notes (a) and (d) of

Table 3.3.1-1 to clarify that the manual enable of the trip is

permissible; and, (4) clarify that the setpoints on Table 3.3.1-1 are

set relative to logarithmic power, not thermal power.

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 TS 3.3.1 does not adversely impact

structure, system, or component design or operation in a manner which

would result in a change in the frequency of occurrence of accident

initiation. SCE has re-analyzed the relevant accidents and established

that accident consequences are not significantly increased by the

proposed changes to the bypass-permissive and enable setpoints. The

reactor trip bypass and automatic enable functions are not accident

initiators. Consequently, the proposed TS change will not significantly

increase the probability of accidents previously evaluated. Therefore,

this amendment request 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.

No new or different accidents result from changing the reactor trip

bypass-permissive and automatic enable setpoints. Introducing an

uncertainty band for the enable setpoints delays the mitigation action

of the reactor trip for the design basis analysis for the events that

credit this trip. The enable setpoint itself does not cause any

accident. Therefore, the amendment request does not create the

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

previously evaluated.

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

a margin of safety.

SCE [Southern California Edison Company] has re-analyzed the

accidents and determined that the consequences of the accidents are

within their acceptance criteria under the proposed amendment so that

the margin of safety that bounds the setpoint in both directions

remains intact. The analyses are relatively insensitive to the reactor

trip automatic enable setpoints, and no significant reduction in the

margins of safety ensues from the relatively minor proposed changes to

the bypass-permissive and enable setpoints, nor from establishing

allowable values for these points.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Main Library, University of

California, Irvine, California 92713.

Attorney for licensee: Douglas K. Porter, Esquire, Southern

California Edison Company, P.O. Box 800, Rosemead, California 91770.

[[Page 71974]]

NRC Project Director: William H. Bateman.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South

Texas Project, Units 1 and 2, Matagorda County, Texas

Date of amendment request: November 23, 1998.

Description of amendment request: The proposed amendment relocates

descriptive design information from Technical Specification 3/4.7.1.1

(Table 3.7-2), regarding orifice sizes for main steam line Code safety

valves, to the Bases section.

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 relocates the orifice size design information

for the main steam line Code safety valves, found in Table 3.7-2, that

does not meet the criteria for inclusion in Technical Specifications as

identified in 10 CFR 50.36(c)(2)(ii). The affected descriptive design

information is not related to any assumed initiators of analyzed events

and is not assumed to mitigate accident or transient events. The

limiting condition for operation for the main steam line Code safety

valves is not altered by the proposed change. The orifice size design

information will be relocated from Table 3.7-2 of Specification 3/

4.7.1.1 to the Bases section for that same Technical Specification and

will be maintained pursuant to 10 CFR 50.59. In addition, surveillance

testing details for this Technical Specification are addressed in

existing surveillance procedures, which are also controlled by 10 CFR

50.59, and subject to the change control provisions imposed by plant

administrative procedures, which endorse applicable regulations and

standards. Therefore, the change does not involve a significant

increase in the probability or consequences of an accident previously

evaluated.

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

different kind of accident from any accident previously evaluated.

The proposed change relocates the orifice size design information

for the main steam line Code safety valves, found in Table 3.7-2, that

does not meet the criteria for inclusion in Technical Specifications as

identified in 10 CFR 50.36(c)(2)(ii). The change does not involve a

physical alteration of the plant (no new or different type of equipment

will be installed) or make changes in the methods governing normal

plant operation. The change will not impose different requirements, and

adequate control of information will be maintained. This change will

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

Therefore, the change does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

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

a margin of safety.

The proposed change relocates the orifice size design information

for the main steam line Code safety valves, found in Table 3.7-2, that

does not meet the criteria for inclusion in Technical Specifications as

identified in 10 CFR 50.36(c)(2)(ii). The change will not reduce a

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

assumptions. In addition, the relocated orifice size design information

remains the same as the existing Technical Specifications. Since any

future changes to this orifice size information (that will be located

in the Bases section) will be evaluated per the requirements of 10 CFR

50.59, there is no reduction in a margin of safety.

The proposed change is also consistent with the Westinghouse Plants

(Improved) Standard Technical Specification, NUREG-1431, approved by

the NRC Staff. Revising the Technical Specification to reflect the

approved content of NUREG-1431 ensures no significant reduction in the

margin of safety. 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 standards of 10 CFR 50.92(c) are

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

request for amendments involves no significant hazards consideration.

Local Public Document Room location: Wharton County Junior College,

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

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

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

NRC Project Director: John N. Hannon.

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

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

Date of amendment request: November 11, 1998.

Brief description of amendments: The proposed amendments revise

core safety limit curves and Overtemperature N-16 reactor trip

setpoints based on analyses of the core configuration and expected

operation for Comanche Peak Steam Electric Station (CPSES) Unit 2,

Cycle 5. The changes apply equally to CPSES Units 1 and 2 licenses

since the Technical Specifications are combined.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

probability or consequences of an accident previously evaluated?

A. Revision to the Unit 2 Core Safety Limits

Analyses of reactor core safety limits are required as part of

reload calculations for each cycle. TU Electric has performed the

analyses of the Unit 2, Cycle 5 core configuration to determine the

reactor core safety limits. The methodologies and safety analysis

values result in new operating curves which, in general, permit plant

operation over a similar range of acceptable conditions. This change

means that if a transient were to occur with the plant operating at the

limits of the new curve, a different temperature and power level might

be attained than if the plant were operating within the bounds of the

old curves. However, since the new curves were developed using NRC

approved methodologies which are wholly consistent with and do not

represent a change in the Technical Specification BASES for safety

limits, all applicable postulated transients will continue to be

properly mitigated. As a result, there will be no significant increase

in the consequences, as determined by accident analyses, of any

accident previously evaluated.

B. Revision to Unit 2 Overtemperature N-16 Reactor Trip Setpoints

As a result of changes discussed, the Overtemperature reactor trip

setpoint has been recalculated. These trip setpoints help ensure that

the core safety limits are protected and that all applicable limits of

the safety analysis are met.

Based on the calculations performed, no significant changes to the

safety

[[Page 71975]]

analysis values for Overtemperature reactor trip setpoint were

required. The f(delta I) trip reset function was revised due to less

top-skewed axial power distributions predicted for this cycle. The

analyses performed show that, using the TU Electric methodologies, all

applicable limits of the safety analysis are met. This setpoint

provides a trip function which allows the mitigation of postulated

accidents and has no impact on accident initiation. Therefore, the

changes in safety analysis values do not involve an increase in the

probability of an accident and, based on satisfying all applicable

safety analysis limits, there is no significant increase in the

consequences of any accident previously evaluated.

In addition, sufficient operating margin has been maintained in the

overtemperature setpoint such that the risk of turbine runbacks or

unnecessary reactor trips due to upper plenum flow anomalies or other

operational transients will be minimized, thereby, reducing potential

challenges to the plant safety systems.

C. Administrative changes to reflect plant nomenclature

Changes to the N-16 trip setpoint equation are for clarification

only to more accurately reflect CPSES plant nomenclature. This change

is administrative in nature and does not increase in the probability or

consequences of an accident previously evaluated.

Summary

The changes in the amendment request apply NRC approved

methodologies to changes in safety analysis values, new core safety

limits and new N-16 setpoint and parameter values to assure that all

applicable safety analysis limits have been met. The potential for an

operational transient to occur has not been affected and there has been

no significant impact on the consequences of any accident previously

evaluated.

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

different kind of accident from any accident previously evaluated?

The proposed changes involve the calculation of new reactor core

safety limits and overtemperature reactor trip setpoint resets. As

such, the changes play an important role in the analysis of postulated

accidents but none of the changes effect plant hardware or the

operation of plant systems in a way that could initiate an accident.

Changes to the N-16 trip setpoint equation are for clarification only

to more accurately reflect CPSES plant nomenclature. Therefore, the

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

kind of accident from any accident previously evaluated.

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

margin of safety?

In reviewing and approving the methods used for safety analyses and

calculations, the NRC has approved the safety analysis limits which

establish the margin of safety to be maintained. While the actual

impact on safety is discussed in response to question 1, the impact on

margin of safety is discussed below:

A. Revision to the Unit 2 Reactor Core Safety Limits

The NRC-approved TU Electric reload analysis methods have been used

to determine new reactor core safety limits. All applicable safety

analysis limits have been met. The methods used are wholly consistent

with Technical Specification BASES 2.1 which is the bases for the

safety limits. In particular, the curves assure that for Unit 2, Cycle

5, the calculated DNBR is no less than the safety analysis limit and

the average enthalpy at the vessel exit is less than the enthalpy of

saturated liquid. The acceptance criteria remains valid and continues

to be satisfied; therefore, no change in a margin of safety occurs.

B. Revision to Unit 2 Overtemperature N-16 Reactor Trip Setpoints

Because the reactor core safety limits for CPSES Unit 2, Cycle 5

are recalculated, the Reactor Trip System instrumentation setpoint

values for the Overtemperature N-16 reactor trip setpoint which protect

the reactor core safety limits must also be recalculated. The

Overtemperature N-16 reactor trip setpoint helps prevent the core and

Reactor Coolant System from exceeding their safety limits during normal

operation and design basis anticipated operational occurrences. The

most relevant design basis analysis in Chapter 15 of the CPSES Final

Safety Analysis Report (FSAR) which is affected by the Overtemperature

reactor trip setpoint is the Uncontrolled Rod Cluster Control Assembly

Bank Withdrawal at Power (FSAR Section 15.4.2). This event has been

analyzed with the new safety analysis value for the Overtemperature

reactor trip setpoint to demonstrate compliance with event specific

acceptance criteria. Because all event acceptance criteria are

satisfied, there is no degradation in a margin of safety.

The nominal Reactor Trip System instrumentation setpoint values for

the Overtemperature N-16 reactor trip setpoint (Technical Specification

Table 2.2-1) are determined based on a statistical combination of all

of the uncertainties in the channels to arrive at a total uncertainty.

The total uncertainty plus additional margin is applied in a

conservative direction to the safety analysis trip setpoint value to

arrive at the nominal and allowable values presented in Technical

Specification Table 2.2-1. Meeting the requirements of Technical

Specification Table 2.2-1 assures that the Overtemperature reactor trip

setpoint assumed in the safety analyses remains valid. The CPSES Unit

2, Cycle 5 Overtemperature reactor trip setpoint is not significantly

different from the previous cycle, and thus provides operational

flexibility to withstand mild transients without initiating automatic

protective actions. Although the value of the f(delta I) trip reset

function setpoint is different, the Reactor Trip System instrumentation

setpoint values for the Overtemperature N-16 reactor trip setpoint are

consistent with the safety analysis assumptions which have been

analytically demonstrated to be adequate to meet the applicable event

acceptance criteria. Thus, there is no reduction in a margin of safety.

Using the NRC approved TU Electric methods, the reactor core safety

limits are determined such that all applicable limits of the safety

analyses are met. Because the applicable event acceptance criteria

continue to be met, there is no significant reduction in the margin of

safety.

C. Administrative changes to reflect plant nomenclature

Changes to the N-16 trip setpoint equation are for clarification

only to more accurately reflect CPSES plant nomenclature. This change

is administrative in nature and has no impact on the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of Texas at

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

19497, Arlington, TX 76019.

Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and

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

NRC Project Director: John N. Hannon.

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, Vermont

Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: December 10, 1998.

[[Page 71976]]

Description of amendment request: The licensee proposed to correct

an error in the technical specifications by changing to the use of

``hydrogen, balance air'' rather than the incorrect ``hydrogen balance

nitrogen'' for calibration of the Augmented Offgass System hydrogen

monitors.

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:

Based on the criteria for defining a significant hazards

consideration in 10CFR50.92, operation of VYNPS in accordance with this

change would not:

(1) Involve a significant increase in the probability or

consequences of an accident previously evaluated, because:

The proposed change is purely administrative in nature--correcting

instrument calibration requirements to conform the Technical

Specification with the instrument manufacturer's recommendations. The

change has no effect on plant hardware, plant design, safety limit

setting, or plant system operation and therefore does not modify or add

any initiating parameters that would significantly increase the

probability or consequences of an accident previously evaluated. This

change to the Technical Specifications is a correction of an error

which occurred when the particular Technical Specification was issued.

The function of this surveillance requirement remains unchanged.

No new modes of operation are introduced by the proposed change

such that adverse consequences would result. Accordingly, the

consequences of previously analyzed accidents are not affected by this

proposed change.

The Augmented Off-Gas (AOG) System hydrogen monitors do not serve a

reactor safety function. In this context, the determination of no

significant hazards consideration defined in 10CFR50.92 is made based

on the ``accident previously evaluated'' being a postulated hydrogen

detonation within the off-gas system downstream of the hydrogen

recombiners. The hydrogen monitors do not mitigate the consequences of

an accident, but rather function to preclude a hydrogen explosion

within the off-gas system. The function of the Augmented Off-Gas System

hydrogen monitors to prevent a hydrogen detonation is not affected by

this change.

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

from any accident previously evaluated, because:

Since this change merely corrects Technical Specification wording

to reflect the actual manufacturer's recommended gas mixture to be used

for calibrating these instruments, no new or different types of

accidents are created. Since the calibration gas mixture has a very low

(approximately 2%) hydrogen concentration, its use does not introduce

the possibility of fires, explosions, or other hazards which might

adversely affect safety-related equipment. Therefore, use of the proper

calibration gas does not create the possibility of a new or different

kind of accident.

This change does not affect the operation of any systems or

components, nor does it involve any potential initiating events that

would create any new or different kind of accident. Therefore, the

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

kind of accident from any previously evaluated for the Vermont Yankee

Nuclear Power Station.

(3) Involve a significant reduction in a margin of safety, because:

This proposed change involving the specification of the correct

calibration gas mixture ensures that the off-gas system hydrogen

monitors are properly calibrated and therefore preserve the margin of

safety in precluding a hydrogen explosion in the off-gas system.

Administratively changing this specification only establishes the

appropriate calibration gas for the actual, installed hydrogen

monitors. Changing the specification to reflect correct practice will

not reduce the margin of safety.

The proposed change does not affect any equipment involved in

potential initiating events or safety limits. Therefore, it is

concluded that the proposed change does not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301.

Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and

Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.

NRC Project Director: Cecil O. Thomas.

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point

Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc

County, Wisconsin

Date of amendment request: July 30, 1998 (TSCR 206).

Description of amendment request: The purpose of the proposed

amendments is to incorporate changes to the Technical Specifications to

more clearly define the requirements for Service Water (SW) System

operability.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration which is presented below:

1. Operation of the Point Beach Nuclear Plant in accordance with

the proposed amendment[s] does not result in a significant increase in

the probability or consequences of any accident previously evaluated.

The Service Water System is primarily a support system for systems

required to be operable for accident mitigation. Portions of the SW

system supplying the containment fan coolers also function as part of

the containment pressure boundary under post accident conditions.

Failures within the SW system are not an initiating condition for any

analyzed accident.

Analyses performed demonstrate that under the Technical

Specifications allowable configurations, the SW system will continue to

perform all required functions. The SW system is capable of supplying

the required cooling water flow to systems required for accident

mitigation. That is, the SW system removes the required heat from the

containment fan coolers and residual heat removal heat exchangers

ensuring containment pressure and temperature profiles following an

accident are as evaluated in the FSAR [final safety analysis report].

This in turn ensures that environmental qualification of equipment

inside containment is maintained and thus function as required post-

accident.

SW system response post accident is within all design limits for

the system. Transient and steady state forces within the system remain

within all design and operability limits thereby maintaining the

integrity of the system inside containment and the integrity of the

containment pressure boundary. Assumptions dependent on containment

pressure profile for containment leakage assumed in the radiological

consequence analyses remain valid.

In addition, removing required heat from containment ensures that

cooling

[[Page 71977]]

of the reactor core is accomplished for long-term accident mitigation.

Therefore, operation of the SW system as proposed will not result

in a significant increase in the probability or consequences of any

accident previously evaluated.

2. Operation of the Point Beach Nuclear Plant in accordance with

the proposed amendments does not result in a new or different kind of

accident from any accident previously evaluated.

The proposed changes do not alter the way in which the SW system

performs its design functions nor the design limits of the system. The

proposed changes do not introduce any new or different normal operation

or accident mitigation functions for the system. Therefore, no new

accident initiators are introduced by the proposed changes. Operation

of SW system as proposed cannot result in a new or different kind of

accident from any accident previously evaluated.

3. Operation of the Point Beach Nuclear Plant in accordance with

the proposed amendments does not result in a significant reduction in a

margin of safety.

Analyses performed in support of the proposed amendments

demonstrate that the SW system continues to perform its function as

assumed and credited in the accident analyses and radiological

consequence analyses performed for the Point Beach Nuclear Plant.

Therefore, the analyses and results are not changed. All analysis

limits remain met. The SW system continues to be operated and responds

within all design limits for the system. Therefore, operation of the

Point Beach Nuclear Plant in accordance with the proposed amendments

cannot result in a reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: The Lester Public Library,

1001 Adams Street, Two Rivers, Wisconsin 54241.

Attorney for licensee: John H. O'Neill, Jr., Shaw, Pittman, Potts,

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

NRC Project Director: Cynthia A. Carpenter.

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point

Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc

County, Wisconsin

Date of amendment request: September 23, 1998 (TSCR 209).

Description of amendment request: The purpose of the proposed

amendments is to remove the test requirements for snubbers from the

Technical Specifications (TS). These requirements are already included

in the Point Beach Nuclear Plant In-Service Inspection Program.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration which is presented below:

1. Operation of the Point Beach Nuclear Plant in accordance with

the proposed amendments will not result in a significant increase in

the probability or consequences of an accident previously evaluated.

These changes do not involve a significant increase in the

probability of an accident previously evaluated because no such

accidents are affected by the proposed revisions to delete TS 15.4.3.

The proposed TS change does not introduce any new accident initiators.

Initiating conditions and assumptions are unchanged and remain as

previously analyzed for accidents in the PBNP Final Safety Analysis

Report. The proposed TS change does not involve any physical changes to

systems or components, nor does it alter the typical manner in which

the systems or components are operated. Therefore, these changes do not

increase the probability of previously evaluated accidents.

As noted above, the snubber testing requirements included in the

ASME/ANSI OM-4 Code are more comprehensive and in general more

conservative than the snubber testing requirements currently contained

in TS 15.4.13.

These changes do not involve a significant increase in the

consequences of an accident or event previously evaluated because the

source term, containment isolation or radiological releases are not

being changed by these proposed revisions. The snubber program ensures

that snubbers function as required, therefore related systems continue

to function as designed and analyzed. Existing system and component

redundancy and operation is not being changed by these proposed

changes. The assumptions used in evaluating the radiological

consequences in the PBNP Final Safety Analysis Report are not

invalidated. Therefore, these changes do not affect the consequences of

previously evaluated accidents.

2. Operation of the Point Beach Nuclear Plant in accordance with

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

different kind of accident from any accident previously evaluated.

These changes do not introduce nor increase the number of failure

mechanisms of a new or different type than those previously evaluated

since there are no physical changes being made to the facility. As

noted above, the snubber testing requirements included in the ASME code

in general are more comprehensive than the snubber testing requirements

currently contained in TS 15.4.13 and provide the requisite level of

assurance of snubber operability. The design and design basis of the

facility remain unchanged. The plant safety analyses remain unchanged.

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

any accident previously evaluated is not introduced.

3. Operation of the Point Beach Nuclear Plant in accordance with

the proposed amendments does not involve a significant reduction in a

margin of safety.

The proposed changes do not involve a significant reduction in the

margin of safety because existing component redundancy is not being

changed by these proposed changes. There are no changes to the initial

conditions contributing to accident severity or consequences, and

safety margins established through the design and facility license

including the Technical Specifications remain unchanged. Therefore,

there are no significant reductions in a margin of safety introduced by

[these] proposed amendment[s].

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: The Lester Public Library,

1001 Adams Street, Two Rivers, Wisconsin 54241.

Attorney for licensee: John H. O'Neill, Jr., Shaw, Pittman, Potts,

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

NRC Project Director: Cynthia A. Carpenter.

[[Page 71978]]

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, Point

Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc

County, Wisconsin

Date of amendment request: October 7, 1998 (TSCR 207).

Description of amendment request: The purpose of the proposed

amendments is to incorporate changes to the Technical Specifications

(TS) to ensure the 4 kV bus undervoltage input to reactor trip is

controlled in accordance with the design and licensing basis for the

facility. One additional administrative change is requested which

removes the footnote related to the definition of Rated Power in TS

15.1.j.

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. Operation of the Point Beach Nuclear Plant [PBNP] in accordance

with the proposed amendments will not create a significant increase in

the probability or consequences of an accident previously evaluated.

The changes proposed ensure the Point Beach Nuclear Plant continues

to be operated in accordance with the design and licensing basis for

the facility.

The first change removes a footnote qualifying the definition of

Rated Power as applied to PBNP Unit 2. This restriction was eliminated

with the replacement of Unit 2 steam generators as approved by

Amendments 173 and 177, dated July 1, 1997. The analyses for those

amendments were performed based on the minimum flow requirements

specified in Technical Specification 15.3.1.G.3. The note should have

been deleted from the Technical Specifications at that time.

Elimination of this note does not result in a change in the operation

of PBNP from that analyzed and approved in Amendments 173 and 177.

Therefore, this change is administrative and cannot result in an

increase in probability or consequences of an accident previously

evaluated.

The second change modifies the Limiting Condition For Operation

[LCO] for the undervoltage reactor trip protection function. This trip

function is the primary protective function credited in the complete

loss of flow event analysis in the Final Safety Analysis Report (FSAR)

Section 14.1.8. As a primary protective function, this trip is required

to be single failure proof as stipulated in proposed IEEE 279-1968

documented in FSAR Section 7.2. This change ensures that this

protective feature is maintained in a condition where single failure

considerations are satisfied. When single failure criteria cannot be

met, appropriate action is stipulated to shutdown the unit placing it

in a condition where the protective function is no longer required.

Therefore, this change ensures PBNP is operated in accordance with its

design and licensing basis and cannot result in an increase in the

probability or consequences of an accident previously evaluated.

2. Operation of the Point Beach Nuclear Plant in accordance with

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

different kind of accident from any accident previously evaluated.

The changes proposed by this request remove a footnote qualifying

the definition of rated power as it applies to PBNP Unit 2 operation,

and modify the LCO related to the undervoltage reactor trip protective

function to ensure this function is maintained as required by the PBNP

design and licensing basis. These changes are in agreement with

approved analyses. These changes do not introduce any new accident

initiators or alter the response of the PBNP Units to previously

analyzed accidents. Therefore, operation of PBNP in accordance with the

proposed changes cannot result in a new or different kind of accident

from any accident previously evaluated.

3. Operation of the Point Beach Nuclear Plant in accordance with

the proposed amendments does not create a significant reduction in a

margin of safety.

Operation of the PBNP in accordance with the proposed amendments is

within the bounds of approved design and licensing basis of the

facility. The design and licensing basis establish appropriate margins

of safety. Since operation of the PBNP remains within the approved

design and licensing basis of the facility, a reduction in a margin of

safety cannot result.

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: The Lester Public Library,

1001 Adams Street, Two Rivers, Wisconsin 54241.

Attorney for licensee: John H. O'Neill, Jr., Shaw, Pittman, Potts,

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

NRC Project Director: Cynthia A. Carpenter.

Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee

Nuclear Power Plant, Kewaunee County, Wisconsin

Date of amendment request: November 18, 1998

Description of amendment request: The proposed amendment would

revise the pressure/temperature (P/T) limits and the low-temperature

overpressure protection (LTOP) requirements in the facility technical

specifications.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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.

Failure of a reactor vessel is not an accident that has been

previously evaluated; design provisions ensure that this is not a

credible event. Since the potential consequences of a reactor vessel

failure are so severe, industry and governmental agencies have worked

together to ensure that failure will not occur. Compliance with 10 CFR

50 Appendix G and H ensures that failure of a reactor vessel will not

occur. The proposed changes do not impact the capability of the reactor

coolant pressure boundary piping (i.e., no change in operating

pressure, materials, seismic loading, etc.) and therefore do not

increase the potential for the occurrence of a LOCA [loss-off-coolant

accident].

The LTOP setpoint, revised enabling temperature, and revised P/T

limits reflected in proposed Figures TS 3.1-1 and TS 3.1-2 ensure that

the Appendix G pressure/temperature limits are not exceeded, and

therefore, ensure that RCS integrity is maintained. The changes do not

modify the reactor coolant system pressure boundary, nor make any

physical changes to the facility design, material, construction

standards, or setpoints. The reactor coolant system full power

operating pressure (2235 psig) is not being changed by this proposed

amendment. The LTOP valve setpoint remains at less than or equal to 500

psig. The LTOP enabling temperature based on Figure

[[Page 71979]]

TS 3.1-2 is 200 deg.F and is consistent with ASME Code Case N-514

guidance of RTNDT + 50 deg.F. The revised enabling

temperature is lower than the 355 deg.F value in the current TS.

However, the allowable combination of Appendix G pressures and

temperatures (refer to the 0 deg.F isothermal cooldown limit) is

greater for the revised limit curves. The combination of greater

allowable Appendix G pressure and temperature limits and lower enabling

temperature produces a larger operating window. A larger operating

window reduces the likelihood of inadvertently lifting the LTOP relief

valve while maneuvering the plant through the knee of the P-T curve

during startup and shutdown. The probability of an LTOP event occurring

is independent of the pressure-temperature limits for the RCS [reactor

coolant system] pressure boundary and enabling temperature. Therefore,

the probability of a[n] LTOP event is not increased.

The revised heatup and cooldown limit curves and LTOP enabling

temperature were developed using test results from unirradiated and/or

irradiated specimens that represent the KNPP [Kewaunee Nuclear Power

Plant] reactor vessel beltline circumferential weld, closure head

flange, and intermediate forging. The circumferential beltline weld and

intermediate forging are the most limiting materials in the reactor

coolant pressure boundary due to the effects of neutron irradiation

which cause the flow properties to increase and the toughness to

decrease. 10 CFR 50, Appendix G states that the metal temperature of

the closure flange regions must exceed the material unirradiated

RTNDT by at least 120 deg.F for normal operation and

90 deg.F for hydrostatic pressure tests and leak tests when the

pressure exceeds 20 percent of the preservice hydrostatic test

pressure. Drop weight and Charpy V-notch testing of IP3571 weld metal

and the intermediate forging material has been performed and used for

derivation of the revised PTS [pressurized thermal shock] assessment,

the proposed Appendix G heatup and cooldown limit curves, and the

corresponding LTOP system enabling temperature. The revised limit

curves and corresponding LTOP enabling temperature have been developed

using accepted engineering practices, methods derived from the ASME

Boiler and Pressure Vessel Code, criteria set forth in NRC Regulatory

Standard Review Plan 5.3.2, and 10 CFR 50.61. Utilization of the

revised heatup and cooldown limit curves and corresponding LTOP

enabling temperature ensures adequate fracture toughness for ferritic

materials of the pressure-retaining components of the reactor coolant

pressure boundary. These limit curves provide adequate margins of

safety during any condition of normal operation, including anticipated

operational occurrences and system hydrostatic tests, and low

temperature overpressure protection (corresponding to isothermal events

during low temperature operations (i.e., less than or equal to

200 deg.F)) thus ensuring the integrity of the reactor coolant pressure

boundary.

The changes do not adversely affect the integrity of the RCS such

that its function in the control of radiological consequences is

affected. Radiological off-site exposures from normal operation and

operational transients, and faults of moderate frequency do not exceed

the guidelines of 10 CFR 100. In addition, the changes do not affect

any fission product barrier. The changes do not degrade or prevent the

response of the LTOP relief valve or other safety-related systems to

previously evaluated accidents. In addition, the changes do not alter

any assumption previously made in the radiological consequence

evaluations nor affect the mitigation of the radiological consequences

of an accident previously evaluated. Therefore, the consequences of an

accident previously evaluated will not be increased.

Thus, operation of KNPP in accordance with the PA does not involve

a significant increase in the probability or consequences of any

accident previously evaluated.

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

from any previously evaluated.

Since the potential consequences of a reactor vessel failure are so

severe, industry and governmental agencies have worked together to

ensure that failure will not occur. Compliance with 10 CFR 50 Appendix

G and H ensures that failure of a reactor vessel will not occur. The

proposed heatup and cooldown limit curves have been constructed by

combining the most conservative pressure-temperature limits derived by

using material properties of the intermediate forging, closure head

flange, and beltline circumferential weld to form a single set of

composite curves. With NRC approval to use Code Case N-588, the

intermediate forging and closure head flange become the controlling

materials for development of the heatup limit curve and the cooldown

limit curves at low temperatures. At high temperatures, the

circumferential weld continues to be limiting for development of the

cooldown limit curves. Use of conservative pressure-temperature limits

derived by using material properties of the intermediate forging,

closure head flange, and beltline circumferential weld to form a single

set of composite curves, does not modify the reactor coolant system

pressure boundary, nor make any physical changes to the LTOP setpoint

or design. Proposed Figures TS 3.1-1 and TS 3.1-2 were prepared in

accordance with regulatory and code requirements and were derived using

more conservative material property basis and more limiting

requirements of neutron exposure projections thru 33 EFPY [effective

full-power years] instead of 20 EFPY.

The revised LTOP system enabling temperature and the proposed

Appendix G pressure temperature limitations were prepared using methods

derived from the ASME Boiler and Pressure Vessel Code and the criteria

set forth in NRC Regulatory Standard Review Plan 5.3.2. The changes do

not cause the initiation of any accident nor create any new credible

limiting failure for safety-related systems and components. The changes

do not result in any event previously deemed incredible being made

credible. As such, it does not create the possibility of an accident

different than previously evaluated.

The changes do not have any adverse effect on the ability of the

safety-related systems to perform their intended safety functions. The

combination of higher allowable Appendix G pressure and temperature

limits and lower enabling temperature produces a larger operating

window. The ASME Section XI, Working Group on Operating Plant Criteria

(WGOPC) has prepared a technical bases document for Code Case N-514.

The technical bases document is contained in Attachment 3 of Reference

1. This technical bases document provides justification for enabling

the LTOP system at temperatures less than 200 deg.F or at coolant

temperatures corresponding to a reactor vessel metal temperature less

than RTNDT + 50 deg.F, whichever is greater.

WGOPC, which has responsibility for Appendix G of Section XI, has

considered the burden and safety impact imposed by the LTOP criteria,

and has developed Code guidelines for determining the LTOP set-point

pressure and the required enabling temperature. These guidelines will

relieve some operational restrictions, yet provide adequate margins

against failure for the reactor vessel. Further, by relieving the

operational restrictions, these guidelines result in a reduced

[[Page 71980]]

potential for activation of pressure relieving devices, thereby

improving plant safety. Thus, a slightly larger operating window at

KNPP is viewed to reduce the likelihood of inadvertently lifting the

LTOP relief valve while maneuvering the plant through the knee of the

P-T curve during startup and shutdown. The new LTOP operating window

(i.e., less than or equal to 200 deg.F) is within the existing

operating band for the residual heat removal system; operating

procedures allow the LTOP system to be placed into service at

<400 deg.F. At KNPP, as long as the LTOP relief valve is operable, the

LTOP system is enabled anytime the residual heat removal system is in

communication with the reactor coolant system.

The proposed changes do not make physical changes to the plant or

create new failure modes. Thus, the PA does not create the possibility

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

evaluated.

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

The proposed Appendix G pressure temperature limitations and LTOP

enabling temperature were prepared using methods derived from the ASME

Boiler and Pressure Vessel Code, including Code Cases N-514 and N-588,

and the criteria set forth in NRC Regulatory Standard Review Plan

5.3.2. Reference 1 to this letter provides information to support NRC

approval to use Code Case N-514 and Code Case N-588 for the KNPP PTS

evaluation, development of the heatup and cooldown limit curves, and

establishment of the LTOP system enabling temperature. These documents

and practices along with the calculational limitations specified in 10

CFR 50.61 are an acceptable method for implementing the requirements of

10 CFR 50 Appendices G and H.

Use of the methodology set forth in the ASME Boiler and Pressure

Vessel Code, NRC Regulatory Standard Review Plan 5.3.2., 10 CFR 50.61,

and 10 CFR 50 Appendices G and H ensures that proper limits and safety

factors are maintained. Thus, the PA does not involve a significant

reduction in the margin of safety.

The revised heatup and cooldown limit curves and LTOP system

enabling temperature were prepared using drop weight and Charpy V-notch

data for the beltline weld, closure head flange, and intermediated

forging material along with practices described herein and methods

derived from the ASME Boiler and Pressure Vessel Code and 10 CFR 50.61.

The safety factors and margins used in the development of the limit

curves and LTOP system enabling temperature meet the criteria set forth

by these documents. Application of low leakage core designs decreases

the rate of shift in transition temperature from ductile to nonductile

behavior. The revised limit curves and LTOP enabling temperature

provide adequate margins of safety during any condition of normal

operation, including anticipated operational occurrences and system

hydrostatic tests, and low temperature overpressure protection

(corresponding to isothermal events during low temperature operations

(i.e., less than or equal to 200 deg.F)). With the preparation of the

revised limit curves in accordance with the latest criteria and

guidance, this PA ensures that proper limits and safety factors are

maintained.

Thus, the PA does not involve a significant reduction in a margin

of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of 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: Cynthia A. Carpenter.

Previously Published Notices of Consideration of Issuance of Amendments

to Facility Operating Licenses, Proposed No Significant Hazards

Consideration Determination, and Opportunity for a Hearing

The following notices were previously published as separate

individual notices. The notice content was the same as above. They were

published as individual notices either because time did not allow the

Commission to wait for this biweekly notice or because the action

involved exigent circumstances. They are repeated here because the

biweekly notice lists all amendments issued or proposed to be issued

involving no significant hazards consideration.

For details, see the individual notice in the Federal Register on

the day and page cited. This notice does not extend the notice period

of the original notice.

Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile Point

Nuclear Station Unit No. 1, Oswego County, New York

Date of application for amendment: May 15, 1998, as supplemented

September 25 and October 13, 1998.

Brief description of amendment: The amendment would revise

Technical Specification 5.5, ``Storage of Unirradiated and Spent Fuel''

to reflect a planned modification to increase the number of fuel

assemblies that can be stored in the spent fuel pool from 2776 to 4086.

Date of publication of individual notice in Federal Register:

November 24, 1998 (63 FR 64973).

Expiration date of individual notice: December 24, 1998.

Local Public Document Room location: Reference and Documents

Department, Penfield Library, State University of New York, Oswego, New

York 13126.

Notice of of Issuance of Amendments to Facility Operating Licenses

During the period since publication of the last biweekly notice,

the Commission has issued the following amendments. The Commission has

determined for each of these amendments that the application complies

with the standards and requirements of the Atomic Energy Act of 1954,

as amended (the Act), and the Commission's rules and regulations. The

Commission has made appropriate findings as required by the Act and the

Commission's rules and regulations in 10 CFR Chapter I, which are set

forth in the license amendment.

Notice of Consideration of Issuance of Amendment to Facility

Operating License, Proposed No Significant Hazards Consideration

Determination, and Opportunity for A Hearing in connection with these

actions was published in the Federal Register as indicated.

Unless otherwise indicated, the Commission has determined that

these amendments satisfy the criteria for categorical exclusion in

accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),

no environmental impact statement or environmental assessment need be

prepared for these amendments. If the Commission has prepared an

environmental assessment under the special circumstances provision in

10 CFR 51.12(b) and has made a determination based on that assessment,

it is so indicated.

For further details with respect to the action see (1) the

applications for amendment, (2) the amendment, and (3)

[[Page 71981]]

the Commission's related letter, Safety Evaluation and/or Environmental

Assessment as indicated. All of these items are available for public

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

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

document rooms for the particular facilities involved.

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,

Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County,

Maryland

Date of application for amendments: October 16, 1998.

Brief description of amendments: The amendments revise Technical

Specification (TS) 3.3.1 ``Reactor Protective System (RPS)

Instrumentation-Operating'' and TS 3.3.2, ``Reactor Protective System

(RPS) Instrumentation-Shutdown,'' to clarify an inconsistency between

the TS wording and the design bases as described in the TS Bases and

the Updated Final Safety Analysis Report. Specifically, the change

replaces the operating bypass input process variable, Thermal Power, in

Footnotes (a), (b), and (d) of Table 3.3.1 and in the Note to Limiting

Condition for Operation 3.3.2 with Nuclear Instrument Power.

Date of issuance: December 8, 1998.

Effective date: As of the date of issuance to be implemented within

30 days.

Amendment Nos.: 229 & 204.

Facility Operating License Nos. DPR-53 and DPR-69: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: October 27, 1998 (63 FR

57320).

The Commission's related evaluation of these amendments is

contained in a Safety Evaluation dated December 8, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Calvert County Library, Prince

Frederick, Maryland 20678.

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

Station, Plymouth County, Massachusetts

Date of application for amendment: April 25, 1996, as supplemented

on September 5, 1996, August 8, 1997, March 26, July 31, and August 24,

1998.

Brief description of amendment: This amendment revises Technical

Specifications (TSs) 3/4.5.F.1, ``Core and Containment Cooling

systems'' to extend the allowed outage time (AOT) for the emergency

diesels, TSs 3.9.B.1 and 3.9.B.4, ``Auxiliary Electrical System'' to

reduce the AOT from 7 days to 3 days and reduce the AOT for the

combination of an EDG and startup transformer or shutdown transformer

from 72 hours to 48 hours, and add Configuration Risk Management

Program in TS 5.5, ``Programs and Manuals'' of Section 5.0

``Administrative Controls''. Various TS pages were re-numbered in

Section 5.0. In addition, TSs 3.9, ``Auxiliary Electrical System,'' and

3.9.A, ``Auxiliary Electrical Equipment,'' have been reformatted to be

consistent with TS 3.9.B approved in a previous amendment. The

associated Bases sections have also been changed to reflect the new

TSs.

Date of issuance: December 11, 1998.

Effective date: As of the date of issuance, to be implemented

within 30 days.

Amendment No.: 179.

Facility Operating License No. DPR-35: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: September 23, 1998 (63

FR 50934).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 11, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts 02360.

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee

Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

Date of application of amendments: July 15, 1997, as supplemented

March 3, April 13, June 16, October 26, and November 5, 1998.

Brief description of amendments: The amendments revised the

Technical Specifications to add new requirements for the main steamline

break instrumentation and resolved issues related to Inspection and

Enforcement Bulletin 80-04.

Date of Issuance: December 7, 1998.

Effective date: As of the date of issuance to be implemented

coincident with implementation of the improved Technical

Specifications.

Amendment Nos.: 234--Unit 1; 234--Unit 2; 233--Unit 3.

Facility Operating License Nos. DPR-38, DPR-47, and DPR-55:

Amendments revised the Technical Specifications.

Date of initial notice in Federal Register: September 24, 1997 (62

FR 50001).

The March 3, April 13, June 16, October 26, and November 5, 1998,

letters provided clarifying information that did not change the scope

of the July 15, 1997, application and the initial proposed no

significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 7, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina.

Duquesne Light Company, et al., Docket No. 50-412, Beaver Valley Power

Station, Unit 2, Shippingport, Pennsylvania

Date of application for amendment: September 24, 1998, as

supplemented November 3, 1998.

Brief description of amendment: This amendment revised technical

specification 3.1.2.8 in two places to change the term ``contained

volume'' to usable volume.'' This change eliminates the potential for a

non-conservative interpretation of the specification values for the

Refueling Water Storage Tank and Boric Acid Storage Tank and thereby

eliminates the need for temporary administrative controls, which have

been used correctly to properly interpret the specification values as

usable volumes.

Date of issuance: December 14, 1998.

Effective date: Effective immediately, to be implemented within 30

days.

Amendment No: 95.

Facility Operating License No. NPF-73. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: November 4, 1998 (63 FR

59591).

The November 3, 1998, letter did not change the initial proposed no

significant hazards consideration determination or expand the amendment

request beyond the scope of the initial notice.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 14, 1998.

No significant hazards consideration comments received: No.

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

663 Franklin Avenue, Aliquippa, PA 15001.

Illinois Power Company, Docket No. 50-461, Clinton Power Station, Unit

1, DeWitt County, Illinois

Date of application for amendment: August 17, 1998.

[[Page 71982]]

Brief description of amendment: The amendment reduces the load at

which diesel generators are tested.

Date of issuance: December 14, 1998.

Effective date: December 14, 1998.

Amendment No.: 118.

Facility Operating License No. NPF-62: The amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: October 7, 1998 (63 FR

53949).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 14, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: The Vespasian Warner Public

Library, 120 West Johnson Street, Clinton, IL 61727.

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

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

Date of application for amendments: August 1, 1997.

Brief description of amendments: The amendments delete a portion of

a technical specifications surveillance test requirement that specifies

that the steam driven auxiliary feedwater pumps be tested ``when the

secondary steam supply pressure is greater than 310 psig.'' This

removes any misunderstanding that the secondary steam pressure must be

just above 310 psig for this test.

Date of issuance: December 10, 1998.

Effective date: December 10, 1998, with full implementation within

45 days.

Amendment Nos.: 225 and 209.

Facility Operating License Nos. DPR-58 and DPR-74: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: December 31, 1997 (62

FR 68308).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 10, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Maud Preston Palenske Memorial

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

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point

Nuclear Station Unit No. 2, Oswego County, New York

Date of application for amendment: February 5, 1998.

Brief description of amendment: This amendment changes the

Technical Specifications to update the terminology and references to 10

CFR 50.55a(f) and (g) consistent with the 1989 edition of Section XI of

the American Society of Mechanical Engineers Boiler and Pressure Vessel

Code, and consistent with the second 10-year interval of the Inservice

Inspections and Inservice Testing Program Plans.

Date of issuance: December 3, 1998.

Effective date: As of the date of issuance to be implemented within

30 days.

Amendment No.: 84

Facility Operating License No. DPR-63: Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: March 11, 1998 (63 FR

11920).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 3, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Reference and Documents

Department, Penfield Library, State University of New York, Oswego, New

York 13126.

PECO Energy Company, Docket Nos. 50-352 and 50-353, Limerick Generating

Station, Units 1 and 2, Montgomery County, Pennsylvania

Date of application for amendments: August 8, 1996, as supplemented

June 30, 1997 and August 26, 1998.

Brief description of amendments: The amendments eliminate the

response time testing requirements for selected sensors and specified

instrument loops for (1) the reactor protection system, (2) the

isolation system, and (3) the emergency core cooling system.

Date of issuance: December 14, 1998.

Effective date: Both units, as of date of issuance, to be

implemented within 30 days.

Amendment Nos.: 132 and 93.

Facility Operating License Nos. NPF-39 and NPF-85: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: November 6, 1996 (61 FR

57489).

The June 30, 1997 and August 26, 1998, letters provided clarifying

information that did not change the initial proposed no significant

hazards consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 14, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, PA 19464.

Power Authority of the State of New York, Docket No. 50-333, James A.

FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of application for amendment: July 10, 1998, as supplemented

October 16, 1998.

Brief description of amendment: The amendment revised Technical

Specification (TS) 3.6/4.6 and associated bases to relocate portions of

the reactor coolant chemistry to the Updated Final Safety Analysis

Report and to applicable plant procedures. Changes to the relocated

requirements will be controlled by the provisions of 10 CFR 50.59.

Date of issuance: December 1, 1998.

Effective date: As of the date of issuance to be implemented within

30 days.

Amendment No.: 247.

Facility Operating License No. DPR-59: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: July 29, 1998 (63 FR

40560).

The October 16, 1998, submittal fell with the scope of, and did not

change, the initial proposed finding of no significant hazards

consideration.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 1, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Reference and Documents

Department, Penfield Library, State University of New York, Oswego, New

York 13126.

Power Authority of the State of New York, Docket No. 50-333, James A.

FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of application for amendment: March 30, 1998, as supplemented

on October 27, 1998.

Brief description of amendment: The amendment revises the

definition of logic system functional tests, and revises test frequency

requirements for certain instrumentation.

Date of issuance: December 11, 1998.

Effective date: As of the date of issuance to be implemented within

30 days.

Amendment No.: 248.

Facility Operating License No. DPR-59: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: April 22, 1998 (63 FR

19978).

The October 27, 1998, supplemental letter provided clarifying

information that did not change the initial proposed no significant

hazards consideration.

[[Page 71983]]

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 11, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Reference and Documents

Department, Penfield Library, State University of New York, Oswego, New

York 13126

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 application for amendments: August 12, 1998, as

supplemented on October 12, 1998. The October 12, 1998, letter provided

clarifying information that did not change the initial proposed no

sigificant hazards consideration determination.

Brief description of amendments: The amendments revise TS 3/

4.6.1.3, ``Containment Air Locks,'' to change the action statements for

an inoperable air lock. The amendments also revise TS Bases 3/4.6.1.2,

``Containment Leakage,'' to correct an editorial error and TS Bases 3/

4.6.1.3, ``Containment Air Locks,'' to provide additional details

regarding the air locks.

Date of issuance: December 2, 1998.

Effective date: December 2, 1998.

Amendment Nos: 215 and 195.

Facility Operating License Nos. DPR-70 and DPR-75: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: September 9, 1998 (63

FR 48265).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 2, 1998

No significant hazards consideration comments received: No.

Local Public Document Room location: Salem Free Public Library, 112

West Broadway, Salem, NJ 08079.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-

364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,

Alabama

Date of amendments request: December 31, 1997, as supplemented by

letter dated September 11, 1998.

Brief Description of amendments: The amendments revised the

Technical Specifications (TSs) to change the intermediate range neutron

flux reactor trip setpoint and allowable value, and delete the

reference to the reactor trip setpoints in TS 3.10.3, ``Special Test

Exceptions--Physics Tests,'' and TS 3.10.4, ``Special Test Exceptions--

Reactor Coolant Loops.''

Date of issuance: December 8, 1998.

Effective date: As of the date of issuance to be implemented within

30 days from the date of issuance.

Amendment Nos.: Unit 1--140; Unit 2--132.

Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise

the Technical Specifications.

Date of initial notice in Federal Register: February 11, 1998 (63

FR 6998).

The September 11, 1998, letter provided clarifying information that

did not change December 31, 1997, application or the initial proposed

no significant hazards consideration determination.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 8, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Houston-Love Memorial Library,

212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama.

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

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

Date of application for amendments: September 20, 1996 (TS 96-09).

Brief description of amendments: The amendments change the

Technical Specifications to clarify the types of work shifts that are

acceptable when considering the requirements to ensure overtime is not

heavily used on a routine basis by unit staff.

Date of issuance: December 7, 1998.

Effective date: As of the date of issuance to be implemented no

later than 45 days after issuance.

Amendment Nos.: 240 and 230.

Facility Operating License Nos. DPR-77 and DPR-79: Amendments

revise the technical specifications.

Date of initial notice in Federal Register: November 4, 1998 (63 FR

59596).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 7, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

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

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

Date of application for amendments: August 22, 1998, as

supplemented on August 27 and October 8, 1998 (TS 96-08). The August

27, 1998, amendment request superseded the original (August 22, 1998)

request in its entirety.

Brief description of amendments: The amendments revise the Sequoyah

Nuclear Plant Technical Specifications by extending the allowed outage

time for the SQN emergency diesel generators from 72 hours to 7 days.

Date of issuance: December 16, 1998.

Effective date: As of the date of issuance to be implemented no

later than 45 days after issuance.

Amendment Nos.: 241 and 231.

Facility Operating License Nos. DPR-77 and DPR-79: Amendments

revise the technical specifications.

Date of initial notice in Federal Register: October 9, 1996 (61 FR

52969), superseded by a second notice on September 9, 1998 (63 FR

48270). The October 8, 1998, letter provided clarifying information

that did not change the initial no significant hazards consideration

determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated December 16, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

Virginia Electric and Power Company, et al., Docket Nos. 50-338 and 50-

339, North Anna Power Station, Units No. 1 and No. 2, Louisa County,

Virginia

Date of application for amendments: July 28, 1998, as supplemented

October 16, 1998. The October 16, 1998, letter was administrative in

nature and did not change the initial no significant hazards

consideration determination.

Brief description of amendments: The amendments revise the

Technical Specifications to change the Emergency Diesel Generator

section to be consistent with station procedures associated with

steady-state conditions.

Date of issuance: December 10, 1998.

Effective date: December 10, 1998.

Amendment Nos.: 216 and 197.

Facility Operating License Nos. NPF-4 and NPF-7: Amendments revised

the Technical Specifications.

Date of initial notice in Federal Register: September 9, 1998 (63

FR 48272).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated December 10, 1998.

No significant hazards consideration comments received: No.

Local Public Document Room location: The Alderman Library, Special

[[Page 71984]]

Collections Department, University of Virginia, Charlottesville,

Virginia 22903-2498.

Dated at Rockville, Maryland, this 23rd day of December 1998.

For the Nuclear Regulatory Commission.

Elinor G. Adensam,

Acting Director, Division of Reactor Projects--III/IV, Office of

Nuclear Reactor Regulation.

[FR Doc. 98-34440 Filed 12-29-98; 8:45 am]

BILLING CODE 7590-01-P