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 February 1, 1997, through February 13, 1997.

The last biweekly notice was published on February 12, 1997 (62 FR

6567).

[[Page 8791]]

Notice of Consideration of Issuance of Amendments to Facility Operating

Licenses, Proposed No Significant Hazards Consideration Determination,

and Opportunity for a Hearing

The Commission has made a proposed determination that the following

amendment requests involve no significant hazards consideration. Under

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

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

Involve a significant increase in the probability or consequences of an

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

or different kind of accident from any accident previously evaluated;

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

for this proposed determination for each amendment request is shown

below.

The Commission is seeking public comments on this proposed

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

publication of this notice will be considered in making any final

determination.

Normally, the Commission will not issue the amendment until the

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

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

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

the Commission may issue the license amendment before the expiration of

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

the amendment involves no significant hazards consideration. The final

determination will consider all public and State comments received

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

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

opportunity for a hearing after issuance. The Commission expects that

the need to take this action will occur very infrequently.

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

Review and Directives Branch, Division of Freedom of Information and

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

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

publication date and page number of this Federal Register notice.

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

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

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

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

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

petitions for leave to intervene is discussed below.

By March 28, 1997, the licensee may file a request for a hearing

with respect to issuance of the amendment to the subject facility

operating license and any person whose interest may be affected by this

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

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

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

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

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

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

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

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

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

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

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

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

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

Secretary or the designated Atomic Safety and Licensing Board will

issue a notice of a hearing or an appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

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

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

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

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

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

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

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

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

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

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

the specificity requirements described above.

Not later than 15 days prior to the first prehearing conference

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

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

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

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

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

brief explanation of the bases of the contention and a concise

statement of the alleged facts or expert opinion which support the

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

contention at the hearing. The petitioner must also provide references

to those specific sources and documents of which the petitioner is

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

facts or expert opinion. Petitioner must provide sufficient information

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

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

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

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

petitioner who fails to file such a supplement which satisfies these

requirements with respect to at least one contention will not be

permitted to participate as a party.

Those permitted to intervene become parties to the proceeding,

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

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

hearing, including the opportunity to present evidence and cross-

examine witnesses.

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

determination on the issue of no significant hazards consideration. The

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

If the final determination is that the amendment request involves

no significant hazards consideration, the Commission may issue the

amendment and make it immediately effective, notwithstanding the

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

of the amendment.

If the final determination is that the amendment request involves a

significant hazards consideration, any hearing held would take place

before the issuance of any amendment.

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

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

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

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

[[Page 8792]]

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

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

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

inform the Commission by a toll-free telephone call to Western Union at

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

operator should be given Datagram Identification Number N1023 and the

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

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

publication date and page number of this Federal Register notice. A

copy of the petition should also be sent to the Office of the General

Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,

and to the attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

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

Station, Plymouth County, Massachusetts.

Date of amendment request: January 30, 1997.

Description of amendment request: The proposed amendment would

change the Updated Final Safety Analysis Report (FSAR) to include the

credit for containment overpressure in the Pilgrim Nuclear Power

Station net positive suction head (NPSH) analysis for the emergency

core cooling pumps.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Will crediting post-LOCA [loss-of-coolant accident] wetwell

airspace pressure in ECCS [emergency core cooling system] analyses

involve a significant increase in the probability or consequences of

an accident previously evaluated?

Chapter 14 of the FSAR contains evaluations of the design basis

accidents, which include the refueling accident, the main steam line

break outside primary containment, the recirculation line break

inside primary containment, and the control rod drop accident. No

increase in the probability of the evaluated accidents will result

from crediting the post-LOCA wetwell airspace pressure because post-

LOCA wetwell airspace pressure does not represent an accident

initiator but is rather a byproduct of the conditions which will

exist in the containment after the pipe break inside containment.

The worst radiological consequences for the Pilgrim plant are

associated with the design basis LOCA which is the double guillotine

failure of the recirculation system piping. The radiological

analysis of this event, contained in FSAR Chapter 14, uses a TID-

14844 source term and assumes a 1.5% per day leakage from the

containment, which is greater than the maximum leakage allowed by

the Technical Specifications. The results of this analysis are

presented in Table 14.5-2 of the FSAR and indicate substantial

margin when compared to 10 CFR Part 100 limits.

The radiological consequences of the design basis accident are

not increased by taking credit for the post-LOCA wetwell airspace

pressure. Assuming containment integrity exists, the mechanism for

increasing the consequences of the accident would be an increased

leakage rate caused by an increase of the average differential

pressure between primary and secondary containment during the

accident response. However, the NPSH analyses performed for Pilgrim,

which credits the post-LOCA wetwell airspace, does not require that

the differential pressure between primary and secondary containment

be maintained above the minimum that exists due to the equilibrium

conditions based on the suppression pool temperature. Specifically,

the wetwell airspace pressure credited in the ECCS pump NPSH

analyses is provided by an increase in wetwell vapor pressure and

air/nitrogen partial pressure in equilibrium with increasing pool

temperature with an accounting for containment initial conditions

and leakage.

By crediting the post-LOCA wetwell airspace pressure in the

calculation of NPSH, no requirement is created to purposely maintain

a higher containment pressure than would otherwise occur; no

requirement is incurred to delay operating containment heat removal

equipment at the highest rate possible; no requirement is incurred

to deliberately continue any condition of high containment pressure

to maintain adequate NPSH; and no requirement is incurred for the

purposeful addition of air/nitrogen into the containment to increase

the available pressure.

Based on these reasons, the probability of accidents previously

evaluated is not increased, and the consequences of the design basis

accident are not increased.

(2) Will crediting post-LOCA wetwell airspace pressure create

the possibility for new or different kinds of accidents?

As stated above, Chapter 14 of the Pilgrim FSAR contains

evaluations of design basis accidents that include the refueling

accident, the main steam line break outside primary containment, the

recirculation line break inside primary containment, and the control

rod drop accident. New or different types of accidents are not

created by crediting the post-LOCA wetwell airspace pressure because

post-LOCA wetwell airspace pressure does not represent an accident

initiator but is rather a byproduct of the conditions which will

exist in the containment after the pipe break inside containment.

Therefore, crediting post-LOCA wetwell airspace pressure does

not create the possibility for new or different kinds of accidents

from those previously analyzed.

(3) Will crediting post-LOCA wetwell airspace pressure in ECCS

NPSH analyses involve a significant reduction in a margin of safety?

The integrity of the primary containment and the operation of

the ECCS systems in combination limit the off-site doses to values

less than those suggested in 10 CFR 100 in the event of a break in

the primary system piping. In order for the ECCS pumps to meet their

performance requirements, the NPSH available to the pumps throughout

the accident response must meet their specific NPSH requirements.

Excess NPSH margin will not improve the performance of the ECCS

pumps because NPSH available must only meet NPSH requirements for

the pump to operate on its pump curve and meet design expectations.

Crediting post-LOCA wetwell airspace pressure in ECCS NPSH

analyses increases the NPSH available to the pumps connected to the

suppression pool but limits the increase in NPSH available

consistent with the bounding leakage assumptions for the containment

system. The amount of post-accident pressure that is utilized in

ECCS NPSH analyses is calculated in a manner such that the pressure

credited represents a conservative lower bound of the pressure

available. Therefore, it is expected that the NPSH margin will

exceed that credited in the NPSH analyses.

Credit for wetwell airspace pressure in NPSH analyses is not

required under all circumstances. If the suction strainers for the

ECCS pumps remain relatively free of post-LOCA debris, adequate NPSH

will be available without credit for the wetwell airspace pressure

provided by the post-LOCA heatup of the air/nitrogen gas in the

containment. If debris accumulates on the pump suction strainers,

the NPSH available to the ECCS pumps will be decreased due to the

head loss caused by the debris. Credit for the post-LOCA wetwell

airspace pressure in the analyses indicates that there is adequate

NPSH margin such that NPSH available will remain above NPSH

required, and ECCS pump performance will meet applicable

requirements. Based on the above discussion, credit for wetwell

airspace pressure in ECCS NPSH analyses 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.

[[Page 8793]]

Therefore, the NRC staff proposes to determine that the amendment

request involves no significant hazards consideration.

Local Public Document Room location: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts 02360.

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

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

NRC Project Director: Patrick D. Milano, Acting.

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

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

North Carolina.

Date of amendments request: November 1, 1996.

Description of amendments request: The amendments would revise the

Technical Specifications (TS) to allow full implementation of the

Boiling Water Reactor Owners Group (BWROG) Enhanced Option 1-A Reactor

Stability Long Term Solution. In Safety Evaluation Reports (SERs)

transmitted to Kevin P. Donovan, Chairman, BWROG, by letters from

Robert C. Jones, Office of Nuclear Reactor Regulation, NRC, dated June

21, 1996, and September 20, 1996, the NRC staff concluded that Enhanced

Option 1-A generic technical specifications described in Topical Report

NEDO-32339, Supplement 4, were acceptable for referencing in license

applications.

The characteristics of a reactor system most important in

determining stability performance are power, core flow and power

distribution. The proposed changes would delete the current limits on

power and flow conditions in the technical specifications associated

with the implementation of the guidance in General Electric Service

Information Letter (SIL) #380, Revision 1 and the power/flow figure

(Figure 3.4.1.1-1), add two new specifications on the fraction of core

boiling boundary (FCBB) and the Period Based Detection System (PBDS)

and relocate certain requirements pertaining to the Average Power Range

Monitors (APRM) to the Core Operating Limits Report (COLR).

The current Technical Specifications for Units 1 and 2 permit

single loop operation (SLO) only for a 12-hour period and there are no

provisions for potential alterations of safety limits or operating

limits because of SLO conditions. Approval of the amendment

applications discussed above would permit SLO operation subject to the

compensatory actions and requirements that address this mode of

operation in the revised Technical Specifications. However, Brunswick

Unit 2's License currently has a condition, 2.C.(5) that states that

the reactor shall not be made critical unless both recirculation loops

are in service. This License Condition also requires the plant to be

placed in the hot shutdown condition within 24 hours if one

recirculation loop becomes out-of-service. The License Condition also

allows one or both recirculation loops to be out-of-service for the

purposes of testing (not to exceed 24 hours). Whereas the License

Condition would permit SLO for up to 24 hours, the current TS limit SLO

to 12 hours. The License Condition was added to permit natural

circulation testing as required by the startup test program but to

preclude long-term SLO or operation in the natural circulation mode.

The startup test program was completed many years ago for Brunswick

Unit 2 and natural circulation operation is no longer allowed. The

License Condition is no longer relevant and if not deleted would negate

the objectives of the proposed license amendments discussed above. The

licensee has submitted proposed license amendments on the same date of

the subject application (i.e., November 1, 1996) to convert the

Brunswick Units 1 and 2 Technical Specifications to the Improved

Standard Technical Specifications (ISTS) consistent with NUREG-1433,

Revision 1, ``Standard Technical Specifications for General Electric

Plants, BWR 4.'' Attachment 6 of the later application was a proposed

revision of the Brunswick Unit 2 License to delete License Condition

2.C.(5). While the Notice of Consideration of Issuance of the ISTS

amendments (62 FR 3719) discussed deletion of License Condition

2.C.(5), the deletion is discussed in this Notice as well, since if the

subject amendment applications are approved, the License Condition

would thwart the considerable effort represented by the subject

amendments to finally resolve the thermal-hydraulic stability issues

for Brunswick Units 1 and 2.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendments do not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed amendments allow the implementation of the Enhanced

Option 1-A (E1A) long term solution to the neutronic/thermal

hydraulic instability issue. Current Technical Specification

restrictions on power and flow conditions, number of operating

recirculation loops and operator actions implemented to reduce the

probability of neutronic/thermal hydraulic instability are

eliminated and new stability control requirements consistent with

NEDO-32339, Supplement 4, are imposed. These requirements include

restrictions on power and flow conditions and actions associated

with the modified APRM flow biased scram and control rod block

functions. These actions include adherence to the boiling boundary

limit stability control prior to entry and during operation in the

region of the power and flow operating domain which is potentially

susceptible to neutronic/thermal hydraulic instability in the

absence of the stability control. In addition, the proposed

amendments require operator actions based upon a new Period Based

Detection System (PBDS). The PBDS is designed to provide alarm

indication that conditions consistent with a significant degradation

in the stability performance of the reactor has occurred and the

potential for imminent onset of neutronic/thermal hydraulic

instability may exist.

The proposed amendments will permit operation in regions of the

power and flow operating domain postulated to be susceptible to

neutron/thermal hydraulic instability (i.e., Restricted and

Monitored Regions). Operation in these regions does not increase the

probability of occurrence of initiators and precursors of previously

analyzed accidents when neutronic/thermal hydraulic instability is

not possible. The proposed amendments also permit the implementation

of the features of the E1A solution which prevent neutronic/thermal

hydraulic instability including pre-emptive reactor scram upon entry

into the region of the power and flow operating domain most

susceptible to neutronic/thermal hydraulic instability (i.e.,

Exclusion Region). Furthermore, the E1A solution requires

implementation of stability control prior to entry into a region of

the power and flow operating domain which is potentially

susceptible, in the absence of stability control, to neutronic/

thermal hydraulic instability (i.e., Restricted Region). The E1A

solution prevents neutronic/thermal hydraulic instability during

operation in regions of the power and flow operating domain

previously excluded from operation and therefore does not

significantly increase the probability of a previously analyzed

accident.

Operation in the regions of the power and flow operating domain

excluded by current Technical Specification 3/4.4.1.1 and Figure

3.4.1.1-1 can occur as a result of anticipated operational

occurrences. The severity of these transients may increase in the

absence of operator actions due to the potential occurrence of

neutronic/thermal hydraulic instability as a result of operation in

these regions. The proposed amendments will permit the

implementation of the E1A long term solution to the stability issue.

Required features of the E1A solution include adherence to a boiling

boundary limit stability control prior to selection by the operator

of APRM flow biased scram and control rod block function setpoints

which

[[Page 8794]]

allow operation in a region of the power and flow operating domain

potentially susceptible, in the absence of the stability control, to

neutronic/thermal hydraulic instability. Upon entry, as a result of

an anticipated operational occurrence, into the region most

susceptible to neutronic/thermal hydraulic instability during

operation with the boiling boundary limit stability control met, the

pre-emptive reactor scram prevents neutronic/thermal hydraulic

instability. Therefore, the consequences of an accident do not

significantly increase while operating with the stability control

met. After exiting the region requiring the stability control to be

met, the setpoints are automatically returned to the values

applicable when anticipated operational occurrences can be initiated

from conditions with the stability control not met. This automatic

actuation of the more conservative setpoints ensures that the pre-

emptive reactor scram will prevent operation as a result of an

anticipated operational occurrence in the region most susceptible to

neutronic/thermal hydraulic instability should the operator not

select the more conservative setpoints appropriate for operation

following exit from the region requiring stability control. These

required features of the E1A solution prevent operation in the

region of the power and flow operating domain most susceptible to

postulated neutronic/thermal hydraulic instability by pre-emptive

reactor scram regardless of how the region was entered. Therefore,

the proposed amendments prevent the occurrence of neutronic/thermal

hydraulic instability as a consequence of an anticipated operational

occurrence and do not significantly increase the consequences of any

previously analyzed accident.

2. The proposed amendments do not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

The proposed amendments eliminate restrictions on power and flow

conditions and impose alternative restrictions which permit the

implementation of the E1A long term stability solution. The current

restrictions on the power and flow conditions do not prevent the

entry into regions of the power and flow operating domain most

susceptible to neutronic/thermal hydraulic instability and therefore

the possibility of neutronic/thermal hydraulic instability exists in

the absence of operator action. The required features of the E1A

solution implement a pre-emptive scram upon entry into the region

most susceptible, without operator action, to neutronic/thermal

hydraulic instability. The accessible operating domain allowed by

the proposed amendments is a subset of the power and flow operating

domain currently allowed. Current initiators and precursors of

accidents and anticipated operational occurrences can not occur with

new or different initial conditions. Therefore, the proposed

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

of accident from that previously evaluated.

Concurrent with the implementation of the proposed amendments, a

modified Flow Control Trip Reference (FCTR) card and a new Period

Based Detection System (PBDS) will be installed as required by the

E1A solution. The function of the FCTR card is to aid the operator

in the identification of entry into regions of the power and flow

operating domain potentially susceptible to neutronic/thermal

hydraulic instability and to initiate a pre-emptive scram upon entry

into the regions most susceptible to neutronic/thermal hydraulic

instability. This is accomplished by altering the values of

setpoints of the APRM flow biased scram and the control rod block

functions generated by the modified FCTR card, which are existing

functions of the current FCTR card. The modified FCTR card design

includes components which may be susceptible to electromagnetic

interference or other environmental effects. The plant specific

environmental conditions (temperature, humidity, pressure, seismic,

and electromagnetic compatibility) have been confirmed to be

enveloped by the PBDS environmental qualification values and will be

confirmed to be enveloped by the E1A FCTR card environmental

qualification values prior to installation. Therefore, the potential

for spurious scrams or common mode failures induced by environmental

effects (e.g., electromagnetic interference) is considered

negligible. The installation of the modified FCTR card will

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

accident from any accident previously evaluated. The function of the

PBDS is to provide the operator with an indication that conditions

consistent with a significant degradation in the stability

performance of the reactor has occurred and the potential for

imminent onset of neutronic/thermal hydraulic instability may exist.

This is accomplished by the installation of a new PBDS card in the

Neutron Monitoring System. The PBDS card takes inputs from

individual local power range monitors and provides displays

indicating alarm and status conditions to the operator in the

control room. These displays can not create the possibility of a new

or different kind of accident from any accident previously

evaluated. The PBDS card design includes components which may be

susceptible to electromagnetic interference or other environmental

effects. The plant specific environmental conditions (temperature,

humidity, pressure, seismic, and electromagnetic compatibility) have

been confirmed to be enveloped by the PBDS environmental

qualification values and will be confirmed to be enveloped by the

E1A FCTR card environmental qualification values prior to

installation. Therefore, the installation of the PBDS card will not

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

any accident previously evaluated.

3. The proposed amendments do not involve a significant

reduction in a margin of safety. The proposed amendments permit the

implementation of the E1A long term solution to the stability issue.

Under certain conditions, existing BWR designs are susceptible to

neutronic/thermal hydraulic instability. General Design Criterion

(GDC) 12 OF 10 CFR 50, Appendix A, requires thermal hydraulic

instability to be prevented by design or be readily and reliably

detected and suppressed. When the design of the reactor system does

not prevent the occurrence of neutronic/thermal hydraulic

instability, instability is an anticipated operational occurrence.

GDC 10 of 10 CFR 50, Appendix A, requires that specified acceptable

fuel design limits not be exceeded during anticipated operational

occurrences.

Analyses performed by the BWROG indicate that neutronic/thermal

hydraulic instability induced power oscillations could result in

conditions exceeding the Minimum Critical Power Ratio (MCPR) Safety

Limit (SL) prior to detection and suppression by the current design

of the Neutron Monitoring System and Reactor Protection System. To

ensure compliance with GDC 12, the BWROG developed Interim

Corrective Actions (ICAs) to enhance the capability of the operator

to readily and reliably detect and suppress neutronic/thermal

hydraulic instability. The BWROG ICAs also provided additional

guidance for monitoring local power range monitors beyond the

requirements of current Technical Specification 3/4.4.1.1 to ensure

adequate margin to the onset of neutronic/thermal hydraulic

instability. Reliance on operator actions to comply with GDC 12 was

accepted on an interim basis by the NRC pending final implementation

of a long term solution to the stability issue.

The modified design of the Reactor Protection System (APRM flow

biased scram) implemented with the E1A solution prevents neutron/

thermal hydraulic instability. The E1A solution also requires

implementation of the stability control prior to entry into a region

of the power and flow operating domain which is potentially

susceptible, in the absence of the stability control, to neutronic/

thermal hydraulic instability. As a result, the margin to the onset

of neutronic/thermal hydraulic instability provided by the existing

Technical Specification requirements and BWROG ICAs recommendations

is not significantly reduced by the implementation of the E1A

solution. The E1A solution assures compliance with GDC 12 by the

prevention of neutronic/thermal hydraulic instability and therefore

precludes neutronic/thermal hydraulic instability from becoming a

credible consequence of an anticipated operational occurrence. The

consequences of anticipated operational occurrences and the margin

to the MCPR SL will not change upon the implementation of the E1A

solution. Therefore, the proposed amendments do not involve a

significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of North Carolina

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

Wilmington, North Carolina 28403-3297.

[[Page 8795]]

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

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

Raleigh, North Carolina 27602.

NRC Project Director: Mark Reinhart (Acting).

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle

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

Date of amendment request: January 20, 1997.

Description of amendment request: The proposed amendments would

relocate the surveillance requirements for selected instrumentation

from the Technical Specifications to licensee controlled documents

because the instrumentation provides indication or an alarm only. The

affected surveillance requirements are: 4.1.3.5.b, ``Control Rod Scram

Accumulators''; 4.5.1.d.2.c, `` Emergency Core Cooling Systems--

Operating''; 4.5.3.1.b, ``ECCS--Suppression Chamber''; and 4.6.2.1.c,

``Containment Systems--Suppression Chamber''. In addition, the proposed

amendments would replace TS SR 4.4.3.2.1, ``Reactor Coolant System

Leakage'' and SR 4.5.1.d.1, ``ECCS--Operating'' with surveillances more

appropriate to the associated LCOs and action statements. Also, the

proposed amendments add an action statement to TS 3.5.1, ``ECCS--

Operating'' regarding pressure of the ADS accumulator backup compressed

gas system bottle, and delete action statements 3.5.3.c, 3.5.3.d,

3.6.2.1.c and 3.6.2.1.d regarding suppression chamber water level

instrumentation.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Involve a significant increase in the probability or

consequences of an accident previously evaluated because:

The proposed change relocates instrumentation requirements,

which provide no post-accident function from the Technical

Specifications to the Bases, UFSAR, procedures, or other plant

controlled documents. These requirements are part of routine

operational monitoring and are not considered in the safety

analysis. The Bases, UFSAR, procedures, and other plant controlled

documents containing the relocated information will be maintained in

accordance with 10 CFR 50.59. In addition to 10 CFR 50.59

provisions, the Technical Specification Bases are subject to the

change control provisions in the Administrative Controls Chapter of

the Technical Specifications. The UFSAR is subject to the change

control provisions of 10 CFR 50.71(e), and plant procedures and

other plant controlled documents are subject to controls imposed by

plant administrative procedures, which endorse applicable

regulations and standards. Since any changes to the Bases, UFSAR,

procedures, or other plant controlled documents will be evaluated

per the requirements of 10 CFR 50.59, no significant increase in the

probability or consequences of an accident previously evaluated will

be allowed. Therefore, this change does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The Reactor Coolant Operational Leakage limits monitoring

surveillance 4.4.3.2.1 has been modified to eliminate procedural

details of what instrumentation/leakage detection systems to use in

verifying limits. The proposed surveillance requires verification

that the reactor coolant system leakage is within limits at the same

frequency as the current surveillance requirement. The reactor

coolant leakage detection systems operability requirements are

controlled by Technical Specification 3/4.4.3.1. Since any changes

to procedures describing the method of monitoring leakage will be

evaluated per the requirements of 10 CFR 50.59, no significant

increase in the probability or consequences of an accident

previously evaluated will be allowed. Therefore, this change does

not involve a significant increase in the probability or

consequences of an accident previously evaluated.

The monitoring action and the surveillance requirements added

for the Automatic Depressurization System (ADS) pneumatic supply

help assure the continued operability of ADS for the mitigation of

accidents involving high reactor vessel pressure and the loss of the

high pressure core spray system. The surveillance frequency is

reasonable for the ADS supply header pressure due to the redundancy

of the instrument nitrogen system, [and] several alarms [that warn]

of system trouble. The ADS accumulator backup compressed gas system

bottle pressure monitoring surveillance frequency and the proposed

action on low bottle pressure is reasonable due to the [presence of

the] ADS accumulator check valves and the [availability of the]

normal ADS supply header. Therefore, this change does not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

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

accident from any accident previously evaluated because:

The proposed change does not involve a physical alteration of

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

or a change in the methods governing normal plant operation. The

proposed change will not impose or eliminate any requirements, and

adequate control of the requirements will be maintained. Thus, these

changes do 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

because:

The proposed change will not reduce a margin of safety because

it has no impact on any safety analysis assumption. In addition, the

requirements to be transposed from the Technical Specifications to

procedures, or other plant controlled documents are the same as the

existing Technical Specifications. Since any future changes to these

requirements in the Bases, UFSAR, procedures, or other plant

controlled documents will be evaluated per the requirements of 10

CFR 50.59, no significant reduction in a margin of safety will be

allowed.

Based on 10 CFR 50.92, the existing requirement for NRC review

and approval of revisions to these requirements proposed for

relocation does not have a specific margin of safety upon which to

evaluate. However, since the proposed change is consistent with the

BWR Standard Technical Specifications, NUREG-1434, approved by the

NRC Staff, revising the Technical Specifications to reflect the

approved level of instrumentation requirements ensures no

significant reduction in the margin of safety.

The Reactor Coolant Operational Leakage limits monitoring

surveillance 4.4.3.2.1 has been modified to eliminate procedural

details of what instrumentation/leakage detection systems to use in

verifying limits. The proposed surveillance requires verification

that the reactor coolant system leakage is within limits at the same

frequency as the current surveillance requirement. The reactor

coolant leakage detection systems operability requirements are

controlled by Technical Specification 3/4.4.3.1. Because there are

no changes to either the reactor coolant leakage detection systems

and the reactor coolant leakage continues to be maintained within

the specified limits, at the required frequency, there is no

reduction in the margin of safety.

The monitoring action and the surveillance requirements added

for the Automatic Depressurization System (ADS) pneumatic supply

help assure the continued operability of ADS for the mitigation of

accidents involving high reactor vessel pressure and the loss of the

high pressure core spray system. This helps assure ADS is maintained

in a ready status. The previous TS SRs only tested the

instrumentation, and did not verify the parameter remained within

limits. Therefore, the margin of safety is not reduced.

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

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

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

requested amendments involve no significant hazards consideration.

Local Public Document Room location: Jacobs Memorial Library,

Illinois Valley Community College, Oglesby, Illinois 61348.

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

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

NRC Project Director: Robert A. Capra.

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

Buren County, Michigan.

[[Page 8796]]

Date of amendment request: January 10, 1996.

Description of amendment request: The proposed amendment would

revise test requirements for the containment emergency escape airlock.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

The following evaluation supports the finding that operation of

the facility in accordance with the proposed change to the Technical

Specifications would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed change does not alter any plant operating

conditions, operating practices, equipment design, equipment

settings, or equipment capabilities. Therefore, operation of the

facility in accordance with the proposed change will not involve an

increase in the probability of an accident. This determination is

made because the full pressure test and the seal contact check

provides reasonable assurance that the Emergency Escape Airlock

doors will act as designed to maintain containment integrity.

Procedures are established to test seal integrity with full pressure

airlock test and to verify the seal contact following the test.

Acceptance criteria are established for each evolution. Failure to

meet the acceptance criteria would result in corrective action to

restore the Emergency Escape Airlock to the intended condition.

The proposed change defines the pressure tests required for the

Emergency Escape Airlock and specifies the method used to restore

the airlock door seals after full pressure testing. Due to the

design of the airlock, the doors must be opened after testing. This

change recognizes the practice of verifying the final integrity of

the airlock by verifying door seal contact. Since the pressure test

does not load the door seals in the same direction as a design basis

accident, this seal contact check provides better assurance that the

door is sealed than alternative pressure tests. The Emergency Escape

Airlock continues to be capable of performing its design function

and the consequences of those accidents previously evaluated will

not increase.

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

from any previously evaluated.

The proposed change does not alter any plant operating

conditions, operating practices, equipment design, equipment

settings, or equipment capabilities. Therefore, operation of the

facility in accordance with the proposed change will 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 change requires testing of the Emergency Escape

Airlock at full pressure (greater than or equal to Pa) rather

than a reduced pressure between-the-seals test. This reduced

pressure test is allowed by the existing Technical Specifications

when the door is opened during periods when containment integrity is

required. The door seal contact check and restoration will provide

assurance that the Emergency Escape Airlock is capable of performing

its design function after the doors are opened during recovery from

full pressure testing. Implementation of these test requirements and

meeting the acceptance criteria will ensure that containment

integrity with respect to the Emergency Escape Airlock will be

maintained. Therefore, there will be no reduction in the margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Van Wylen Library, Hope

College, Holland, Michigan 49423.

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

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

NRC Project Director: John N. Hannon.

Duke Power Company, Docket Nos. 50-269, 270 and 50-287, Oconee

Nuclear Station, Units 1, 2 and 3, Oconee County, South Carolina.

Date of amendment request: February 5, 1997 (TSC 96-11)

Description of amendment request: The proposed changes would

reflect replacement of the existing nuclear instrumentation with an

enhanced wide range nuclear instrumentation system that provides more

channels and continuous coverage from the source to above the power

range. As a result: (1) The various references to Intermediate Range of

nuclear instrumentation would be eliminated and replaced with reference

to Wide Range instrumentation; (2) the minimum number of operable

Source and Wide Range Nuclear Instrumentation channels that are

available and that are required to be operable in Table 3.5.1-1 would

be increased; (3) the minimum power level specified in Note (c) of

Table 3.5.1-1 would be changed from 10-10 amps on the intermediate

range instrument channels to 4 x 10-4% rated power on the wide

range instrument channels; and (4) entries that specify the Wide Range

Nuclear Instrumentation, the number of Required Operable Channels,

reference to a new Action Statement, and Applicability would be added

to Table 3.5.6-1.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Will the change involve a significant increase in the

probability or consequences of an accident previously evaluated?

No. The proposed amendment to the Oconee Technical

Specifications is associated with the implementation of an enhanced

nuclear instrumentation system. The new Gamma Metrics system

provides twice the number of channels of neutron detectors for use

during both normal plant operations and post-accident monitoring.

The proposed change will make Oconee's Technical Specifications

consistent with a nuclear instrumentation system that meets the

reliability and redundancy requirements of Regulatory Guide 1.97.

Additionally, the new Technical Specifications will be more

conservative in terms of stating the minimum number of operable

channels required, since there are now a greater number of redundant

channels available. Assuring that the nuclear instrumentation at

Oconee is more reliable and more redundant, does not affect the

probability of an occurrence of an accident, since this system is a

monitoring system and not an accident initiator. However, these

characteristics (increased reliability and redundancy) could provide

additional capability to deal with the consequences of post-accident

situations.

(2) Will the change create the possibility of a new or different

kind of accident from any [kind of accident] previously evaluated?

No. The proposed amendment to Oconee Technical Specifications

involves the implementation of an enhanced nuclear instrumentation

system. By implementing a nuclear instrumentation system that meets

the provisions of Regulatory Guide 1.97, Oconee's ability for

neutron monitoring is enhanced during normal operations and post-

accident recovery. The Source Range nuclear instrumentation system

is utilized for monitoring purposes only, while the Wide Range

provides a control rod withdrawal interlock based on high startup

rate. The new Gamma Metrics detectors have been shown to be more

reliable, accurate, and redundant than Oconee's original detectors.

Therefore, changing the Oconee Technical Specifications to be

consistent with the current nuclear instrumentation arrangement, as

proposed in this amendment request, has no effect on the possibility

of any type of accident: new, different, or previously evaluated.

(3) Will the change involve a significant reduction in a margin

of safety?

No. Margin of safety is associated with confidence in the

ability to maintain the fission product barriers (i.e., fuel and

fuel cladding, Reactor Coolant System pressure boundary, and

containment structure) to limit the level of radiation dose to the

public. The proposed Technical Specifications amendment will

establish operability requirements for an enhanced nuclear

instrumentation system at Oconee. By

[[Page 8797]]

implementing a more reliable and redundant nuclear instrumentation

system, Oconee's post-accident monitoring capability is enhanced.

Therefore, the ability to protect the public from radiation dose is

further assured, and no reduction in any existing margin of safety

will occur.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Oconee County Library, 501

West South Broad Street, Walhalla, South Carolina 29691.

Attorney for licensee: J. Michael McGarry, III, Winston and Strawn,

1200 17th Street, NW., Washington, DC 20036.

NRC Project Director: Herbert N. Berkow.

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

Nuclear Station, Unit 1, Claiborne County, Mississippi.

Date of amendment request: October 22, 1996.

Description of amendment request: The proposed amendment would

revise Figure 3.4.11-1, ``Minimum Reactor Vessel Metal Temperature vs.

Reactor Vessel Pressure,'' in Limiting Condition for Operation 3.4.11,

``RCS [Reactor Coolant System] Pressure and Temperature (P/T) Limits,''

of the Technical Specifications (TSs). The existing curve is valid only

up to 10 Effective Full Power Years (EFPYs) and would be revised to be

valid up to 32 EFPYs.

The proposed curves, pages 1 through 5 of Figure 3.4.11-1, have

been drawn for five different EFPY periods: 16, 20, 24, 28 and 32.

There are two sets of curves attached to the licensee's application.

The first set of curves (Attachment 3) would replace the existing curve

in TS Figure 3.4.11-1. The second set of curves (Attachment 4) are

duplicates of the Attachment 3 curves except that these curves also

contain detailed information used in development of the curves and

would be included in the next update of the Updated Final Safety

Analysis Report (UFSAR) for information.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, in its application for the proposed amendment, which is

presented below:

(A) The proposed change does not significantly increase the

probability or consequences of an accident previously evaluated.

Regulatory Guide 1.99, Revision 2 is currently used to prepare

the pressure-temperature limit curves and is inherently conservative

for Boiling Water Reactors (BWRs). [Grand Gulf Unit 1 is a BWR.] The

proposed Technical Specification Figure 3.4.11-1 was prepared in

accordance with the requirements of 10CFR50 [10 CFR Part 50],

Appendix G [(Fracture Toughness Requirements)], and using NRC

approved methodology outlined in NRC Regulatory Guide 1.99, Revision

2, ``Radiation Embrittlement of Reactor Vessel Materials.''

Operation of the plant within the limitations of the proposed figure

will ensure that the Requirements of 10CFR50 [10 CFR Part 50],

Appendix G are met up to and including 32 Effective Full Power Years

(EFPY) of operation. The proposed changes assure that the existing

safety limits are not exceeded due to changing Reactor Vessel

conditions by continued incorporation of the effect of neutron

radiation embrittlement of vessel materials into the proposed

curves.

The curves have also been editorially enhanced by removal of

phrases used for validation of the curves. Having the phrases on the

TS (Technical Specification) curves distracts from the intended

purpose which is to maintain operation of the reactor to the right

of the curves. Operators, in performance of their job function, do

not need this information to comply with TS Limiting Condition for

Operation (LCO) 3.4.11. This change also revises the curve labeling

consistent with the terminology used in Table 1 of 10CFR50 [10 CFR

Part 50], Appendix G. These enhancements and revisions have no

impact on the operation of the plant since they are editorial in

nature and do not change the technical content of the curves.

Therefore, the proposed change does not significantly increase

the probability or consequences of an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The pressure-temperature curves are controlled by the Technical

Specifications and are determined using the conservative methodology

in NRC Regulatory Guide 1.99, Revision 2, ``Radiation Embrittlement

of Reactor Vessel Materials.'' The proposed pressure-temperature

limit curves are inherently conservative, therefore, the possibility

of failure of the reactor vessel is not increased. The proposed

curves establish new periods of applicability (16, 20, 24, 28, and

32 EFPY) for the current pressure-temperature limitations based on

NRC methodology in Regulatory Guide 1.99 and actual fluence

measurements. These limitations are appropriate up to and including

32 EFPY exposure and operation of the plant within the figure's

limitations will ensure that the requirements of 10CFR50 [10 CFR

Part 50], Appendix G are met for that time frame. No physical plant

modifications or new operating configurations result from these

changes. These changes do not adversely affect the design or

operation of any system or component important to safety, rather

they establish limits to assure that operations remain within

acceptable safety boundaries.

The curves have also been editorially enhanced by removal of

phrases used for validation of the curves. Having the phrases on the

TS curves distracts from the intended purpose which is to maintain

operation of the reactor to the right of the curves. Operators, in

performance of their job function, do not need this information to

comply with TS Limiting Condition for Operation (LCO) 3.4.11. This

change also revises the curve labeling consistent with the

terminology used in Table 1 of 10CFR50 [10 CFR Part 50], Appendix G.

These enhancements and revisions have no impact on the operation of

the plant since they are editorial in nature and do not change the

technical content of the curves.

Therefore, the proposed change does not create the possibility

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

evaluated.

(C) The proposed change does not involve a significant reduction

in a margin of safety.

The proposed curves were developed using the methodology of

Regulatory Guide 1.99, Revision 2, ``Radiation Embrittlement of

Reactor Vessel Materials.'' This methodology includes an allowance

for margin that is to be included in the upper-bound values of the

adjusted reference temperature (ART). The proposed changes maintain

the existing margins of safety by modifying the operating limits

based on the most limiting of the actual reference temperature

shifts. These new limits consider the most limiting pressure vessel

material. The revised analysis demonstrates that the existing

Technical Specification [TS] pressure-temperature limit curves are

applicable for periods of 16, 20, 24, 28, and 32 EFPY. Using the

methodology in NRC Regulatory Guide 1.99 Revision 2 and fluence

based on actual exposure provides for additional conservatism, and

therefore [,] further assures the existence of current margins of

safety. The proposed pressure-temperature limit curves are

inherently conservative and provide sufficient margin to ensure the

integrity of the reactor vessel.

The curves have also been editorially enhanced by removal of

phrases used for validation of the curves. Having the phrases on the

TS curves distracts from the intended purpose which is to maintain

operation of the reactor to the right of the curves. Operators, in

performance of their job function, do not need this information to

comply with TS Limiting Condition for Operation (LCO) 3.4.11. This

change also revises the curve labeling consistent with the

terminology used in Table 1 of 10CFR50 [10 CFR Part 50], Appendix G.

These enhancements and revisions have no impact on the operation of

the plant since they are editorial in nature and do not change the

technical content of the curves.

Continuing commitment to the methodology contained in NRC

Regulatory Guide 1.99, Rev. 2, will ensure that the most limiting

plate or beltline weld material will be utilized in the

determination of the pressure-temperature limits for any future

curve changes.

[[Page 8798]]

Therefore, the proposed change does not result in a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room Location: Judge George W. Armstrong

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

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

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

NRC Project Director: William D. Beckner.

Entergy Gulf States, Inc., Cajun Electric Power Cooperative, and

Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit

1, West Feliciana Parish, Louisiana.

Date of amendment request: January 10, 1997.

Description of amendment request: The proposed amendment would

revise the Technical Specifications (TSs) for reactor pressure vessel

pressure and temperature (P-T) limits to replace the curves for 2

effective full power years (EFPY) with curves for 12 EFPY. The P-T

curves are used for heatup, cooldown, and inservice leak and

hydrostatic testing.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of an accident previously evaluated.

Pressure-temperature (P-T) limits (RBS Technical Specifications

Figure 3.4.11-1) are imposed on the reactor coolant system to ensure

that adequate safety margins against nonductile or rapidly

propagating failure exist during normal operation, anticipated

operational occurrences, and system hydrostatic tests. The P-T

limits are related to the nil-ductility reference temperature,

RTNDT, as described in ASME Section III, Appendix G. Changes in

the fracture toughness properties of [Reactor Pressure Vessel] RPV

beltline materials, resulting from the neutron irradiation and the

thermal environment, are monitored by a surveillance program in

compliance with the requirements of 10 CFR [Part] 50, Appendix H.

The effect of neutron fluence on the nil-ductility reference

temperature of pressure vessel steel is predicted by methods given

in Regulatory Guide [RG] 1.99, Rev. 2.

The revised P-T limits of this amendment request were

established based on adjusted reference temperatures developed in

accordance with the procedures prescribed in Reg. Guide [RG] 1.99,

Rev. 2, Regulatory Position C.1. Calculation of adjusted reference

temperature by these procedures includes a margin term to ensure

conservative, upper-bound values are used for the calculation of the

P-T limits. Stress intensity factors used to compute the pressures

were calculated in accordance with, and include the required safety

factors given in ASME Section III, Appendix G. The limits

established by the lower portion of the P-T curves, which cover the

discontinuity (non-beltline) regions of the vessel (e.g., flanges,

nozzles, etc.), were retained throughout this current analysis. The

limits established by the lower portion of these curves do not

change as they are not affected significantly by the neutron

fluence.

This change is not related to any accidents previously

evaluated. The proposed change will provide for approved P-T limit

curves which are valid through 12 EFPY. This change will not affect

any Safety Limits, Power Distribution Limits, or Limiting Conditions

for Operation. The proposed change will not affect reactor pressure

vessel [RPV] performance as no physical changes are involved and RBS

vessel P-T limits will remain conservative in accordance with Reg.

Guide [RG] 1.99, Rev. 2 and ASME Section III, Appendix G

requirements. The proposed change will not cause the reactor

pressure vessel [RPV] or interfacing systems to be operated outside

of their design or testing limits. Also, the proposed change will

not alter any assumptions previously made in evaluating the

radiological consequences of accidents. The proposed change ensures

that adequate margins against brittle fracture of the vessel are

maintained through 12 EFPY of reactor operations. Therefore, the

probability or consequences of accidents previously evaluated will

not be increased by the proposed change.

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

or different kind of accident from any accident previously

evaluated.

The proposed change is a revision of Technical Specification

Figure 3.4.11-1 to show P-T limit curves valid through 12 EFPY. The

revised P-T limits have been established in accordance with

applicable NRC regulations and the ASME Code. This proposed change

does not involve a modification of the design of plant structures,

systems, or components. The proposed change will not impact the

manner in which the plant is operated as plant operating and testing

procedures will not be affected by the change. The proposed change

will not degrade the reliability of structures, systems, or

components important to safety (ITS) as equipment protection

features will not be deleted or modified, equipment redundancy or

independence will not be reduced, supporting system performance will

not be downgraded, the frequency of operation of ITS equipment will

not be imposed. No new accident types or failure modes will be

introduced as a result of the proposed change. Therefore, the

proposed change does not create the possibility of a new or

different kind of accident from that previously evaluated.

3. The proposed changes do not involve a significant reduction

in a margin of safety.

As stated in the River Bend SER, ``Appendices G and H of 10 CFR

50 describe the conditions that require pressure-temperature [P-T]

limits and provide the general bases for these limits. These

appendices specifically require that pressure-temperature [P-T]

limits must provide safety margins at least as great as those

recommended in the ASME Code, Section III, Appendix G. * * * Until

the results from the reactor vessel surveillance program become

available, the staff will use RG 1.99, Revision 1 [now Revision 2]

to predict the amount of neutron irradiation damage. * * * The use

of operating limits based on these criteria--as defined by

applicable regulations, codes, and standards--will provide

reasonable assurance that nonductile or rapidly propagating failure

will not occur, and will constitute an acceptable basis for

satisfying the applicable requirements of GDC 31.''

Bases for RBS Technical Specification 3.4.11 states: ``The P/T

[P-T] limits are not derived from Design Basis Accident (DBA)

analyses. They are prescribed during normal operation to avoid

encountering pressure, temperature, and temperature rate of change

conditions that might cause undetected flaws to propagate and cause

nonductile failure of the RCPB, a condition that is unanalyzed. * *

* Since the P/T [P-T] limits are not derived from any DBA, there are

no acceptance limits related to the P/T [P-T] limits. Rather, the P/

T [P-T] limits are acceptance limits themselves since they preclude

operation in an unanalyzed condition.''

This amendment request proposes P-T limit curves which will be

valid through 12 EFPY. The proposed P-T limits were established

based on adjusted reference temperatures for vessel beltline

material calculated in accordance with Regulatory Position 1 of Reg.

Guide [RG] 1.99, Rev. 2 and pressures calculated in accordance with

ASME Section III, Appendix G requirements. Required margins and

safety factors were included to ensure that conservative, upper-

bound values were used in calculation of the P-T limits. The

proposed change will not affect any Safety Limits, Power

Distribution Limits, or Limiting Conditions for Operation. The

proposed change does not represent a change in initial conditions,

or in a system response time, or in any other parameter affecting

the course of an accident analysis supporting the Bases of any

Technical Specification. The proposed P-T limits provide adequate

safety margins against brittle failure of the reactor vessel through

12 EFPY of power operations. For these reasons, the proposed changes

do not involve a reduction in any margins of safety.

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

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

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

amendment request involves no significant hazards consideration.

[[Page 8799]]

Local Public Document Room location: Government Documents

Department, Louisiana State University, Baton Rouge, LA 70803.

Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,

1400 L Street, NW., Washington, DC 20005.

NRC Project Director: William D. Beckner.

Entergy Gulf States, Inc., Cajun Electric Power Cooperative, and

Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit

1, West Feliciana Parish, Louisiana.

Date of amendment request: January 20, 1997.

Description of amendment request: The proposed amendment would

revise the Technical Specifications (TSs) to allow the use of flow

control spectral shift strategies to increase cycle energy; an

estimated additional 30 days at full power. The request is based on a

General Electric (GE) Maximum Extended Load Line Limit (MELLL) analysis

for the River Bend Station.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed changes do not significantly increase the

probability or consequences of an accident previously evaluated.

Abnormal operational transients or accidents analyzed in the SAR

have been examined for any impact caused by MELLL operation. The

limiting abnormal operation transients, including the Generator Load

Rejection with No Bypass (LRNBP) event and the Feedwater Controller

Failure (FWCF) maximum demand event, have been evaluated in detail.

The LOCA [Loss-of-Coolant Accident], Fuel Loading Error (FLE), rod

drop accident, rod withdrawal error, and the Anticipated Transient

Without Scram (ATWS) analyses have also been evaluated for the

effects of MELLL operation. The flow and power dependent [Minimum

Critical Power Ratio] MCPR curves for off-rated and rated conditions

and the [Maximum Average Planar Linear Heat Generation Rate] MAPLHGR

criteria establish limits on power operation. These limits ensure

that the core is operated within the assumptions and initial

conditions of the transient or accident analyses. Operation within

these limits will ensure that the consequences of a transient or

accident remain within the acceptable limits of the analyses.

The [Average Power Range Monitor] APRM scram in the Technical

Specifications [TSs] and affected rod block setpoints are revised to

ensure that operation remains within the analyzed MELLL region. This

restriction ensures the consequences of abnormal operation and

accidents are acceptable. The probability of an accident is not

affected by the proposed Technical Specification [TS] changes since

no systems or equipment which could initiate an accident are

affected. Therefore, the proposed changes do not significantly

increase the probability or consequences of any previously evaluated

accident.

2. The request does not create the possibility of occurrence of

a new or different kind of accident from any accident previously

evaluated.

Operation in the MELLL domain expands the current power/flow

along the 121% rod line to 100% power at 75% rated core flow and

improves flexibility and capacity factor. Abnormal operation

transients or accidents have been evaluated and the most limiting

cases have been analyzed for applicability for operation in the

MELLL region. The proposed Technical Specification [TS] changes

prohibit power operation outside the MELLL region and do not

constitute or require any system or equipment changes that might

create an accident of a different type then previously evaluated.

The MAPLHGR, the power and flow dependent MCPR and [Liner Heat

Generation Rate] LHGR and the revised Technical Specifications [TSs]

will continue to assure that plant operation is consistent with the

assumptions, initial conditions and assumed power distribution and

therefore will not create a new type of accident. The proposed

Technical Specification [TS] changes do not introduce any new modes

of plant operation nor involve new system interactions. Therefore,

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

different kind of accident from any previous analyzed.

3. The request does not involve a significant reduction in a

margin of safety.

The proposed Technical Specifications [TSs] prohibit power

operation outside the allowable MELLL region. The transients and

accidents described in the SAR are evaluated for operation in the

MELLL region. NEDC-32611, ``MELLL Analysis for River Bend Station

Reload 6 Cycle 7,'' shows that the OLMCPR for operation in the MELLL

region is bounded by the OLMCPR established for current conditions

(100% power/107% flow). The thermal limits MCPR and LHGR curves and

the MAPLHGR limits establish limits on power operation and thereby

ensure that the core is operated within the assumptions and initial

conditions of the transient and accident analyses.

As demonstrated in the analysis provided in Attachment 4, [the

proposed amendment request] operation within these limits, using the

MCPR limits, LHGH limits and MAPLHGR criteria, will ensure that the

margin of safety will be maintained to the same level described in

the Technical Specifications Bases and the SAR and the consequences

of the postulated transient or accidents are not increased. The MCPR

safety limit, mechanical performance limits and overpressure limit

are not exceeded during any transient or postulated accident.

Therefore, the proposed Technical Specifications [TSs] to allow

operation in the MELLL region do not involve a significant reduction

in margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Government Documents

Department, Louisiana State University, Baton Rouge, LA 70803.

Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,

1400 L Street, N.W., Washington, D.C. 20005.

NRC Project Director: William D. Beckner.

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

Atomic Power Station, Lincoln County, Maine.

Date of amendment request: February 7, 1997.

Description of amendment request: The proposed amendment would

modify Technical Specification 3.12 to require both 115 kV incoming

lines to be operable when the reactor is critical; allow continued

operations for up to 72 hours with one 115 kV incoming line inoperable;

allow continued operations for up to 24 hours with both 115 kV incoming

lines inoperable; apply the increased operability requirements

described above to another affected remedial action; incorporate minor

editorial changes to uniformly apply the usage of the term

``operable;'' and change the basis section to be consistent with the

proposed changes.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change does not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The proposed changes to Specification 3.12.B do not involve a

physical change to the plant or the maintenance of the plant. The

proposed changes increase the operating requirements associated with

the operability of the 115 kV incoming lines beyond that currently

required by Technical Specifications. For those accidents previously

evaluated, the more restrictive operability requirements associated

with maintaining both 115 kV incoming lines operable and the more

restrictive remedial action times result in increased assurance that

station service power will be available when required. This

increased availability will be achieved because elective maintenance

on the offsite power system will be significantly restricted and the

restoration of inoperable 115 kV incoming lines will be treated with

greater urgency. The increased

[[Page 8800]]

assurance of availability will result in a decrease in the

probability or consequences of these postulated accidents.

However, the more restrictive remedial action times decrease the

restoration period and consequently increase the possibility that

successful restoration may not be achieved, given an outage of the

115 kV power system. A unit shutdown without offsite power would

then be commenced. This evolution would involve a unit shutdown

without the availability of equipment such as the reactor coolant

pumps, condensate pumps and main feedwater pumps. Although none of

these components are credited as available for the mitigation of the

consequences of accidents previously evaluated, the probability of

the occurrence of certain accidents is increased without them.

Although the combination of these considerations could involve

an increase in the probability of accidents previously evaluated,

the increase would not be significant due to the low probability of

independent failures or common cause failures of both of the 115 kV

incoming lines. There is no increase in the consequences of any

accident previously evaluated as a result of these proposed

Technical Specification changes. The proposed Technical

Specification changes are consistent with the Standard Technical

Specifications approved by the NRC. The proposed changes, therefore,

will not involve a significant increase in the probability or

consequences of an accident previously evaluated.

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

or different kind of accident from any previously evaluated.

The proposed Technical Specification change does not involve a

change to the physical plant or to the physical configuration of the

offsite power system. The effect of the proposed change will be to

increase the availability of the offsite power system when required.

In addition, the proposed change will increase the possibility of a

unit shutdown without offsite power operable. However, the accidents

previously evaluated assume a simultaneous loss of offsite power,

design basis accident and worst case single failure as part of the

design basis. The proposed changes do not result in the creation of

a unique operating condition or a configuration that has not been

previously evaluated. Therefore, the proposed change does not create

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

accident previously evaluated.

3. The proposed change does not involve a significant reduction

in the margin of safety.

This proposed change modifies Technical Specification 3.12 to be

consistent with the Standard Technical Specifications. The proposed

Technical Specification change maintains the current margin of

safety which is based upon supplying power to engineered safeguards.

Adequate sources of power remain available for the operation of the

engineered safeguards equipment. Therefore, the proposed change

would not involve a significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Wiscasset Public Library, High

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

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

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

NRC Project Director: Patrick D. Milano, Acting.

Northeast Nuclear Energy Company, et al., Docket Nos. 50-245, 50-

336, and 50-423, Millstone Nuclear Power Station, Unit Nos. 1, 2, and

3, New London, Connecticut.

Date of amendment request: February 3, 1997.

Description of amendment request: The licensee has proposed to

revise Section 6, ``Administrative Controls,'' of the Millstone Unit

Nos. 1, 2, and 3 Technical Specifications to reflect organizational

changes that have been implemented in the Nuclear Division.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

consideration] because the changes would not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

No design basis accidents are affected by these proposed

changes. The proposed changes are administrative in nature and are

being proposed to reflect the organizational changes which become

effective on February 3, 1997. The unit level responsibilities of

the Executive Vice President--Nuclear are assigned to the Officers

for the individual Millstone units. The site level responsibilities

of the Executive Vice President--Nuclear are shared by the Senior

Vice President and CNO [Chief Nuclear Officer]--Millstone and the

President and Chief Executive Officer. The changes to the SORC [Site

Operations Review Committee] and the three unit[s'] PORC [Plant

Operations Review Committee] reflect changes in job function or job

position titles only.

No safety systems are adversely affected by the proposed

changes, and no failure modes are associated with the changes.

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

from any accident previously evaluated.

Because there are no changes in the way plants are operated due

to this administrative change, the potential for an unanalyzed

accident is not created. There is no impact on plant response, and

no new failure modes are introduced. These proposed administrative

and editorial changes have no impact on safety limits or design

basis accidents, and they have no potential to create a new or

unanalyzed event. The changes to the SORC and the three unit[s']

PORC reflect changes in job function or job position titles only.

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

The changes do not directly affect any protective boundaries nor

do they impact the safety limits for the protective boundaries.

These proposed changes are administrative and editorial in nature.

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

changes do not reduce the margin of safety provided by the PORC and

the SORC review and approval of changes to the operations of the

Millstone Unit Nos. 1, 2, and 3.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center,

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

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

Rope Ferry Road, Waterford, CT 06385.

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

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

CT 06141-0270.

NRC Deputy Director: Phillip F. McKee.

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: February 5, 1996.

Description of amendment request: The amendment would delete a

clause from Technical Specification 4.0.5.a. Specifically, this change

would delete the clause ``(g), except where specific written relief has

been granted by the Commission pursuant to 10 CFR Part 50, Section

50.55a(g)(6)(i).'' The amendment would also make the appropriate

changes to the Bases section. In addition, NNECO made changes to Bases

Section 3/4.7.7 and 3/4.7.8 to add design basis information and provide

clarification of system design and operation.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

[[Page 8801]]

consideration, which is presented below:

Pursuant to 10 CFR 50.92, NNECO has reviewed the proposed

changes to Technical Specification 4.0.5a and Bases Section 3/4.4.10

and has concluded that the changes do not involve a significant

hazards consideration (SHC). The basis for this conclusion is that

the three criteria of 10 CFR 50.92(c) are not compromised. The

proposed changes do not involve an SHC because the changes would

not:

1. Involve a significant increase in the probability or

consequences of an accident previously evaluated.

The proposed changes would remove the wording ``* * * (g),

except where specific written relief has been granted by the

Commission pursuant to 10 CFR Part 50, Section 50.55a(g)(6)(i).''

The Inservice Inspection and Testing Programs are described in the

technical specifications pursuant to 10 CFR 50.55a. In addition, the

proposed changes, in accordance with NUREG-1431 and NUREG-1482,

would provide relief to the ASME Code requirement in the interim

between the time of submittal of a relief request until the NRC has

issued a safety evaluation and granted the relief. The changes being

proposed are administrative in nature and do not affect assumptions

contained in plant safety analyses, the physical design and/or

operation of the plant, nor do they affect any technical

specification that preserves safety analysis assumptions. Any relief

from the approved ASME Section XI Code requirements will require a

10 CFR 50.59 evaluation to ensure no technical specification changes

or unreviewed safety questions exist. Therefore, operation of the

facility in accordance with the proposed changes would not affect

the probability or consequences of an accident previously analyzed.

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

from any previously evaluated.

The proposed changes would remove the wording ``* * * (g),

except where specific written relief has been granted by the

Commission pursuant to 10 CFR Part 50, Section 50.55a(g)(6)(i).''

The Inservice Inspection and Testing Programs are described in the

technical specifications pursuant to 10 CFR 50.55a. In addition, the

proposed changes, in accordance with NUREG-1431 and NUREG-1482,

would provide relief to the ASME Code requirement in the interim

between the time of submittal of a relief request until the NRC has

issued a safety evaluation and granted relief. The changes being

proposed are administrative in nature and will not change the

physical plant or the modes of operation defined in the facility

license. The changes do not involve the addition or modification of

equipment nor do they alter the design or operation of plant

systems. Any relief from the approved ASME Section XI Code

requirements will require a 10 CFR 50.59 evaluation to ensure no

technical specification changes or unreviewed safety questions

exist. Therefore, operation of the facility in accordance with the

proposed changes would 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 changes would remove the wording ``* * * (g),

except where specific written relief has been granted by the

Commission pursuant to 10 CFR Part 50, Section 50.55a(g)(6)(i).''

The Inservice Inspection and Testing Programs are described in the

technical specifications pursuant to 10 CFR 50.55a. In addition, the

proposed changes, in accordance with NUREG-1431 and NUREG-1482,

would provide relief to the ASME Code requirement in the interim

between the time of submittal of a relief request until the NRC has

issued a safety evaluation and granted relief. The changes being

proposed are administrative in nature and will not alter the bases

for assurance that safety-related activities are performed correctly

or the basis for any technical specification that is related to the

establishment or maintenance of a safety margin. Any relief from the

approved ASME Section XI Code requirements will require a 10 CFR

50.59 evaluation to ensure no technical specification changes or

unreviewed safety questions exist. Therefore, operation of the

facility in accordance with the proposed changes would not involve a

significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: 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,

CT 06141-0270.

NRC Deputy Director: Phillip F. McKee.

United States Department of Commerce, National Institute of

Standards and Technology, Docket No. 50-184, NIST (formerly known as

National Bureau of Standards) Test Reactor or NBSR.

Date of amendment request: January 17, 1997.

Description of amendment request: The National Institute of

Standards and Technology (NIST) is planning to change the name of the

Reactor Radiation Division to the NIST Center for Neutron Research to

be headed by a Director. The requested amendment involves a name change

only. All functions, responsibilities, and personnel remain the same.

The Technical Specification references to the ``Chief, Reactor

Radiation Division'' will be changed to Director, NIST Center for

Neutron Research in Sections 7.1, 7.2, and 7.3. The Organization Chart

in Figure 7.1 will also reflect this change. The Technical

Specification references to the ``Reactor Radiation Division'' will be

changed to ``NIST Center for Neutron Research'' in Section 7.2.

Basis for proposed no significant hazards consideration

determination: The Commission has provided standards for determining

whether a significant hazards consideration exists (10 CFR 50.92(c)). A

proposed amendment would not: (1) Involve a significant increase in the

probability or consequences of an accident previously evaluated; or (2)

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

accident previously evaluated; or (3) involve a significant reduction

in a margin of safety.

The change being proposed is a change in the title of the

organization and the title of the head of the organization that directs

the operation of the reactor. As noted previously, all functions,

responsibilities and personnel remain the same. The staff agrees with

the licensee's no significant hazards consideration and finds that the

mere title changes render a negative response to the three criteria

outlined in 10 CFR 50.92(c).

Local Public Document Room location: N/A.

Attorney for licensee: N/A

NRC Project Director: Seymour H. Weiss.

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

Vermont Yankee Nuclear Power Station, Vernon, Vermont.

Date of amendment request: December 10, 1996.

Description of amendment request: The proposed amendment would move

fire protection requirements from the Vermont Yankee Technical

Specifications to the Fire Protection Plan and the final safety

analysis report (FSAR), in accordance with the guidance in NRC Generic

Letters 86-10 and 88-12.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment will not involve a significant

increase in the probability or consequences of an accident

previously evaluated:

[[Page 8802]]

The proposed changes are administrative in nature and are

consistent with the guidance provided in NRC Generic Letters 86-10

and 88-12. These changes do not affect the initial conditions or

precursors assumed in the FSAR safety analyses. These proposed

changes also do not decrease the effectiveness of equipment relied

upon to mitigate the previously evaluated accidents. Programmatic

controls will continue to assure that fire protection program

changes do not reduce the effectiveness of the program to achieve

and maintain safe shutdown in the event of a fire.

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

new or different kind of accident from an accident previously

evaluated:

The proposed changes do not modify any plant equipment, there is

no reduction in fire protection requirements, there is no change in

operating procedure and surveillance requirements and no reduction

in administrative control or equipment reliability. Therefore,

implementation of the proposed change will not affect the design

function or configuration of any component, introduce any new

operating scenarios, failure modes or accident initiators.

3. The proposed amendment will not involve a significant

reduction in a margin of safety:

The proposed amendment does not involve a reduction to the Fire

Protection Program. The fire protection requirements are simply

being relocated to other controlled documents. There are no

equipment modifications being proposed, only the location of fire

protection requirements, which is administrative in nature.

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: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301.

Attorney for licensee: R. K. Gad, III, Ropes and Gray, One

International Place, Boston, MA 02110-2624.

NRC Project Director: Patrick D. Milano, Acting Director.

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.

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

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

Carolina.

Date of amendment request: January 10, 1997.

Brief description of amendment: The proposed change would revise

Technical Specification 4.8.1.1.2 to clarify pressure testing

requirements for the isolable and non-isolable portions of the diesel

fuel oil piping.

Date of publication of individual notice in Federal Register:

February 5, 1997 (62 FR 5490).

Expiration date of individual notice: March 6, 1997.

Local Public Document Room location: Cameron Village Regional

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

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

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

County, Minnesota.

Date of amendment request: November 6, 1996.

Description of amendment request: The proposed amendments would

revise the Technical Specifications governing the cooling water system.

The changes are proposed to improve plant operation based on

operational experience with the vertical motor-driven cooling water

pump. The changes are also proposed to incorporate information gathered

by the licensee during its self-assessment Service Water System

Operational Performance Inspection (SWSOPI) completed in late 1995.

Date of individual notice in the Federal Register: January 29, 1997

(62 FR 4338).

Expiration date of individual notice: February 28, 1997.

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401.

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

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

County, Minnesota.

Date of amendment requests: January 29, 1997.

Description of amendment requests: The proposed amendments would

change the Bases for Technical Specifications and the licensing basis

for the Operating Licenses relating to the cooling water system

emergency intake line flow capacity. The licensee determined through

testing that the emergency intake line flow capacity was less than the

design value stated in the Updated Final Safety Analysis Report. The

proposed changes reflect the use of operator actions to control cooling

water system flow following a seismic event. The proposed changes also

reclassify the intake canal for use during a seismic event, which would

be an additional source of cooling water during a seismic event.

Date of individual notice in the Federal Register: February 7, 1997

(62 FR 5857).

Expiration date of individual notice: March 10, 1997. NSHC

comments: February 24, 1997.

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

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

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401.

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

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

New Jersey Date of amendment request: January 31, 1997.

Brief description of amendment request: The amendment would make

changes to Technical Specification (TS) 3.4.3, ``Relief Valves,'' for

Salem Unit 1, and TS 3.4.5, ``Relief Valves,'' for Salem Unit 2, to

ensure that the automatic capability of the power operated relief

valves to relieve pressure is maintained when these valves are isolated

by closure of the block valves.

Date of publication of individual notice in Federal Register:

February 7, 1997 (62 FR 5861).

Expiration date of individual notice: March 10, 1997.

Local Public Document Room location: Salem Free Public Library, 112

West Broadway, Salem, NJ 08079.

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: October 18, 1996.

Description of amendments request: Amend Technical Specifications

to permanently incorporate new requirements associated with steam

generator tube inspections and repair. The requirements provide

alternate

[[Page 8803]]

steam generator tube plugging criteria at the tube support plate

intersections.

Date of publication of individual notice in the Federal Register:

February 11, 1997 (62 FR 6276).

Expiration date of individual notice: March 13, 1997.

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

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

Point Beach Nuclear Power Plant, Unit Nos. 1 and 2, Town of Two Creeks:

Manitowoc County, Wisconsin.

Date of amendment requests: September 19, 1996, as supplemented

November 18, 1996, and revised January 13 and January 27, 1997.

Description of amendment requests: The proposed amendments would

change Technical Specification requirements related to the low

temperature overpressure protection (LTOP) system. Specifically, the

reactor coolant system (RCS) temperature below which LTOP is required

to be enabled and the temperature below which one high pressure safety

injection pump is required to be rendered inoperable would be changed

from less than 275 degrees Fahrenheit to less than 355 degrees

Fahrenheit. Additionally, the restriction of ``less than the minimum

pressurization temperature for the inservice pressure test as defined

in Figure 15.3.1-1'' would be deleted and the specific temperature

limit of less than 355 degrees Fahrenheit would be specified. The

setpoint for the pressurizer power-operated relief valves (PORVs) would

be changed from less than or equal to 425 pounds per square inch gage

(psig) to less than or equal to 440 psig to allow for instrument

inaccuracies and increased margin allowed by the use of American

Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code

Case N-514. These modified requirements for LTOP ensure that RCS

materials meet the requirements of Title 10 of the Code of Federal

Regulations, Sec. 50.60, ``Acceptance Criteria for Fracture Prevention

Measures for Lightwater Nuclear Power Reactors for Normal Operation,''

(10 CFR 50.60) in accordance with 10 CFR Part 50, Appendices G and H,

and in accordance with the exemption granted on January 27, 1997, which

allows the use of ASME Code Case N-514 as an acceptable alternative.

Finally, editorial changes would be made to rename the ``Overpressure

Mitigating System'' to the ``Low Temperature Overpressure Protection

System.'' The September 19, 1996, application was previously noticed in

the Federal Register on October 1, 1996 (61 FR 51308).

Date of individual notice in the Federal Register: February 4, 1997

(62 FR 5256).

Expiration date of individual notice: March 6, 1997. NSHC comments

February 19, 1997.

Local Public Document Room location: Joseph P. Mann Library, 1516

Sixteenth Street, Two Rivers, Wisconsin 54241.

Notice 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) 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: November 26, 1996.

Brief description of amendments: The amendments adopt Option B of

10 CFR Part 50, Appendix J to require Type B and Type C containment

leakage testing to be performed on a performance-based testing

schedule.

Date of issuance: February 11, 1997.

Effective date: As of the date of issuance to be implemented within

30 days.

Amendment Nos.: 219 and 196.

Facility Operating License Nos. DPR-53 and DPR-69: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: January 2, 1997 (62 FR

123).

The Commission's related evaluation of these amendments is

contained in a Safety Evaluation dated February 11, 1997.

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

December 23, 1996.

Brief description of amendment: The amendment will revise the

definition of Operable-Operability, revise Technical Specifications

(TSs) and associated Bases Section for TS 3.9.B.2 and 3.9.B.3,

``Auxiliary Electrical System,'' TS 3.4.B.1, ``Standby Liquid Control

System,'' TSs 3.7.b.1.a, c, and e, and 3.7.b.2.a, c, and e, ``Standby

Gas Treatment System and Control Room High Efficiency Air Filtration

System,'' and TSs. 4.5.F.1, ``Core and Containment Cooling Systems,''

and delete TS 3.7.b.1.f, ``Standby Gas Treatment System and Control

Room High Efficiency Air Filtration System.''

Date of issuance: February 10, 1997.

Effective date: February 10, 1997.

Amendment No.: 170.

Facility Operating License No. DPR-35: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: June 19, 1996 (61 FR

31172).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated February 10, 1997.

No significant hazards consideration comments received: No.

Local Public Document Room location: Plymouth Public Library, 11

North Street, Plymouth, Massachusetts 02360.

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

[[Page 8804]]

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

Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos.

1 and 2, Will County, Illinois.

Date of application for amendments: August 2, 1996.

Brief description of amendments: The amendments eliminate License

Condition 2.C.(16) from Facility Operating License NPF-37; License

Condition 2.C.(5) from Facility Operating License NPF-66; License

Condition 2.C.(6) from Facility Operating License NPF-72 and License

Condition 2.C.(5) from Facility Operating License NPF-77 that require

the licensee to conduct additional corrosion testing of sleeved steam

generator tubes.

Date of issuance: February 12, 1997.

Effective date: Immediately, to be implemented within 30 days.

Amendment Nos.: 85 to NPF-37, 85 to NPF-66, 77 to NPF-72, and 77 to

NPF-77.

Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:

The amendments revise the licenses.

Date of initial notice in Federal Register: September 25, 1996 (61

FR 50340).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated February 12, 1997.

No significant hazards consideration comments received: No.

Local Public Document Room location: For Byron, the Byron Public

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

for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street,

Wilmington, Illinois 60481.

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

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

Date of application for amendments: November 26, 1996, as

supplemented December 17, 1996

Brief description of amendments: The amendments revise Technical

Specification 3.8.2.1 to allow a one-time change to replace the

existing 125-volt AT&T high specific gravity round cell battery banks

with the conventional low specific gravity cell battery banks.

Date of issuance: February 7, 1997.

Effective date: As of the date of issuance to be implemented within

30 days.

Amendment Nos.: 172 and 154.

Facility Operating License Nos. NPF-9 and NPF-17: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: December 13, 1996 (61

FR 65605).

The December 17, 1996, letter did not change the scope of the

November 26, 1996, 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 February 7, 1997.

No significant hazards consideration comments received: No.

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

University of North Carolina at Charlotte, 9201 University City

Boulevard, Charlotte, North Carolina 28223-0001.

Entergy Gulf States, Inc., Cajun Electric Power Cooperative, and

Entergy Operations, Inc., Docket No. 50-458, River Bend Station, Unit

1, West Feliciana Parish, Louisiana.

Date of amendment request: August 29, 1996.

Brief description of amendment: The amendment revises the Technical

Requirements Manual (TRM) to change the reactor pressure vessel

surveillance capsule withdrawal schedule for the River Bend Station.

The first capsule will be withdrawn at 10.4 effective full power years

(EFPY) rather than at 6 EFPY.

Date of issuance: February 13, 1997.

Effective date: February 13, 1997.

Amendment No.: 92.

Facility Operating License No. NPF-47. The amendment revised the

Technical Requirements Manual.

Date of initial notice in Federal Register: October 23, 1996 (61 FR

55034) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated February 13, 1997.

No significant hazards consideration comments received. No.

Local Public Document Room location: Government Documents

Department, Louisiana State University, Baton Rouge, LA 70803.

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

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

Date of amendment request: June 27, 1996.

Brief description of amendment: The amendment modifies TS 3/

4.3.3.6, ``Accident Monitoring Instrumentation,'' to reflect the

Combution Engineering improved Standard Technical Specification (STS)

approved and issued as NUREG-1432. This amendment revises the TS to

include Accident Monitoring Instrumentation recommended in Regulatory

Guide (RG) 1.97, ``Instrumentation for Light-Water-

Cooled Nuclear Plants to Assess Plant Conditions During and Following

an Accident,'' Revision 3.

Date of issuance: February 12, 1997.

Effective date: February 12, 1997, to be implemented within 90

days.

Amendment No.: 122.

Facility Operating License No. NPF-38. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: July 3, 1996 (61 FR

40017).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated February 12, 1997.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of New Orleans

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

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

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

Date of amendment request: July 25, 1996, as supplemented by letter

dated January 27, 1997.

Brief description of amendment: The amendment changes the Appendix

A Technical Specifications by modifying TS 3/4.7.4, ``Ultimate Heat

Sink,'' to incorporate more restrictive fan operability requirements

and lower the maximum allowed basin temperature.

Date of issuance: February 13, 1997.

Effective date: February 13, 1997.

Amendment No.: 123.

Facility Operating License No. NPF-38. Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: November 19, 1996 (61

FR 58903).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated February 13, 1997.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of New Orleans

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

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-

389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida.

Date of application for amendments: October 30, 1996.

Brief description of amendments: These amendments revise the St.

Lucie Technical Specifications to remove inconsistencies between the

definition of Core Alterations and the Applicability, Action and

Surveillance requirements of two specifications relating to water level

and containment

[[Page 8805]]

isolation systems during refueling operations.

Date of Issuance: February 10, 1997.

Effective Date: February 10, 1997.

Amendment Nos.: 148 and 87.

Facility Operating License Nos. DPR-67 and NPF-16: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: December 4, 1996 (61 FR

64386).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated February 10, 1997.

No significant hazards consideration comments received: No.

Local Public Document Room location: Indian River Junior College

Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-

389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida.

Date of application for amendment: October 28, 1996.

Brief description of amendment: The amendments consist of changes

to the Technical Specifications (TS) in response to your applications,

both dated October 28, 1996, regarding containment leakage tests and

removal of certain component lists from the TS.

Date of Issuance: February 10, 1997.

Effective Date: February 10, 1997.

Amendment Nos.: 149 and 88.

Facility Operating License No. NPF-16: Amendments revised the

Technical Specifications.

Date of initial notice in Federal Register: (61 FR 64386) December

4, 1996. The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated February 10, 1997.

No significant hazards consideration comments received: No.

Local Public Document Room location: Indian River Junior College

Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.

Florida Power and Light Company, Docket Nos. 50-250 and 50-251,

Turkey Point Plant Units 3 and 4, Dade County, Florida.

Date of application for amendments: December 17, 1996.

Brief description of amendments: Revision to Technical

Specification (TS) 4.4.10 regarding reactor coolant pump flywheel

inspection intervals.

Date of issuance: February 11, 1997.

Effective date: February 11, 1997.

Amendment Nos.: 193 and 187.

Facility Operating Licenses Nos. DPR-31 and DPR-41: Amendments

revise the Technical Specifications.

Date of initial notice in Federal Register: January 10, 1997 (62 FR

1476). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated February 11, 1997.

No significant hazards consideration comments received: No.

Local Public Document Room location: Florida International

University, University Park, Miami, Florida 33199.

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: July 16, 1996.

Brief description of amendment: The amendment changes the Technical

Specifications to permit the use of 10 CFR Part 50, Appendix J, Option

B, Performance-Based Containment Leakage Rate Testing in accordance

with the implementation guidance in NRC's Regulatory Guide 1.163 dated

September 1995.

Date of issuance: February 10, 1997.

Effective date: As of the date of issuance to be implemented within

30 days.

Amendment No.: 159.

Facility Operating License No. DPR-63: Amendment revises the

Technical Specifications.

Date of initial notice in Federal Register: October 9, 1996 (61 FR

52965). The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated February 10, 1997.

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.

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

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

County, Minnesota.

Date of application for amendments: August 15, 1996.

Brief description of amendments: The amendments revise the

containment cooling systems limiting conditions for operation technical

specifications to bring them into conformance with recently completed

system analyses by no longer permitting both containment spray pumps to

be inoperable at the same time.

Date of issuance: February 10, 1997.

Effective date: February 10, 1997, with full implementation within

30 days.

Amendment Nos.: 125 and 117.

Facility Operating License Nos. DPR-42 and DPR-60: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: December 4, 1996 (61 FR

64388).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated February 10, 1997.

No significant hazards consideration comments received: No.

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401.

Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,

Limerick Generating Station, Units 1 and 2, Montgomery County,

Pennsylvania.

Date of application for amendments: November 25, 1996.

Brief description of amendments: These amendments revise the

wording in TS Section 4.8.1.1.2.e.2 and the associated TS Bases Section

3/4.8, to remove the specific reference to the Residual Heat Removal

(RHR) pump motor and its corresponding kW rating value, and replace it

with wording consistent with that specified in the Improved TS (i.e.,

NUREG-1433, Revision 1, ``Standard Technical Specifications General

Electric Plants,'' dated April 1995).

Date of issuance: February 4, 1997.

Effective date: Both units, as of date of issuance, to be

implemented within 30 days.

Amendment Nos.: 121 and 85.

Facility Operating License Nos. NPF-39 and NPF-85. The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: December 18, 1996 (61

FR 66716).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated February 4, 1997.

No significant hazards consideration comments received: No.

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, PA 19464.

Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,

Limerick Generating Station, Units 1 and 2, Montgomery County,

Pennsylvania.

Date of application for amendments: September 27, 1996.

Brief description of amendments: These amendments increase the

reactor enclosure secondary containment maximum inleakage rate, and

also impact secondary containment drawdown time and system flow rate

assumptions, thereby, affecting charcoal filter bed efficiency and post

accident dose analysis.

[[Page 8806]]

Date of issuance: February 11, 1997.

Effective date: Both units, as of the date of issuance, to be

implemented within 30 days.

Amendment Nos.: 122 and 86.

Facility Operating License Nos. NPF-39 and NPF-85. The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: December 4, 1996 (61 FR

64392).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated February 11, 1997.

No significant hazards consideration comments received: No.

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, PA 19464.

Philadelphia Electric Company, Docket No. 50-353, Limerick

GeneratingStation, Unit 2, Montgomery County, Pennsylvania.

Date of application for amendment: December 6, 1996, as

supplemented by letters dated January 15, and 28, 1997.

Brief description of amendment: This amendment modifies Technical

Specification (TS) Section 2.1 and its associated TS Bases to reflect

the change in the Minimum Critical Power Ratio safety limit due to the

use of GE13 fuel product line and the cycle-specific analysis performed

by General Electric Company (GE), for LGS, Unit 2, Cycle 5.

Date of issuance: February 12, 1997.

Effective date: As of date of issuance, to be implemented within 30

days.

Amendment No.: 87.

Facility Operating License No. NPF-85. This amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: December 23, 1996 (61

FR 67582).

The January 15, and 28, 1997, letters provided clarifying

information that did not change the initial proposed no significant

hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated February 12, 1997.

No significant hazards consideration comments received: No.

Local Public Document Room location: Pottstown Public Library, 500

High Street, Pottstown, PA 19464.

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: June 10, 1996, as supplemented

June 24, July 1, August 13, September 20, and October 17, 1996.

Brief description of amendments: The amendments change Technical

Specifications 3/4.3.3.1, ``Radiation Monitoring Instrumentation,'' and

3/4.7.6, ``Control Room Emergency Air Conditioning System,'' to reflect

a control room design in which the common Unit 1 and Unit 2 control

room envelope is supplied by 2 one hundred percent capable Control Room

Emergency Air Conditioning System trains.

Date of issuance: February 6, 1997.

Effective date: Both units, as of date of issuance, to be

implemented within 30 days.

Amendment Nos.: 190 and 173.

Facility Operating License Nos. DPR-70 and DPR-75. The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: June 24, 1996 (61 FR

32468) The June 24, July 1, August 13, September 20, and October 17,

1996, letters provided clarifying information that did not change the

initial proposed no significant hazards consideration determination nor

expand the scope of the initial submittal as described in the initial

notice.

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated February 6, 1997.

No significant hazards consideration comments received: No.

Local Public Document Room location: Salem Free Public Library, 112

West Broadway, Salem, NJ 08079.

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: May 31, 1996, as supplemented

December 23, 1996.

Brief description of amendments: The amendments change the

Technical Specification to (1) Revise the reactor vessel level

indication system action statements, (2) revise the channel calibration

definition, and (3) delete a requirement to install la jumper in the

auxiliary feedwater actuation logic.

Date of issuance: February 6, 1997.

Effective date: Both units, as of its date of issuance, to be

implemented within 60 days.

Amendment Nos.: 191 and 174.

Facility Operating License Nos. DPR-70 and DPR-75. The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: June 17, 1996 (61 FR

30641).

The December 23, 1996, letter 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 February 6, 1997.

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: November 15, 1996.

Brief Description of amendments: The amendments replace Containment

Systems TS 3.6.2.2 for the Spray Additive System, with a new Emergency

Core Cooling Systems (ECCS) TS 3.5.6 for the ECCS Recirculation Fluid

pH Control System.

Date of issuance: February 3, 1997.

Effective date: As of the date of issuance to be implemented prior

to Mode 4 for Unit 1 following the spring 1997 refueling outage; for

Unit 2 following the spring 1998 refueling outage.

Amendment Nos.: 123 and 118.

Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise

the Technical Specifications.

Date of initial notice in Federal Register: December 18, 1996 (61

FR 66718).

The Commission's related evaluation of the amendments is contained

in a Safety Evaluation dated February 3, 1997.

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 36302.

Toledo Edison Company, Centerior Service Company, and The Cleveland

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

Power Station, Unit No. 1, Ottawa County, Ohio.

Date of application for amendment: August 7, 1996.

Brief description of amendment: The amendment revises Technical

Specification (TS) 1.0, ``Definitions,'' by defining a refueling

interval to be [less than or equal to] 730 days; and revises TS 3/4.0,

``Applicability,'' TS 3/4.6.2.1, ``Containment Systems--

Depressurization and Cooling Systems--Containment Spray System,'' and

TS 3/4.6.3.1, ``Containment Systems--

[[Page 8807]]

Containment Isolation Valves,'' to reflect performing surveillance

tests during a refueling interval rather than every 18 months.

Date of issuance: February 10, 1997.

Effective date: February 10, 1997, to be implemented not later than

120 days after issuance.

Amendment No.: 213.

Facility Operating License No. NPF-3: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: October 9, 1996 (61 FR

52970).

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated February 10, 1997.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of Toledo, William

Carlson Library, Government Documents Collection, 2801 West Bancroft

Avenue, Toledo, Ohio 43606.

Toledo Edison Company, Centerior Service Company, and The Cleveland

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

Power Station, Unit No. 1, Ottawa County, Ohio.

Date of application for amendment: September 12, 1996.

Brief description of amendment: The amendment revised Technical

Specifications (TS) 3/4.1.3.4, ``Reactivity Control Systems--Rod Drop

Time,'' and TS 3/4.5.2, ``Emergency Core Cooling Systems--Tavg [greater

than or equal to] 280 deg.F,'' to change the surveillance test interval

from every 18 months to each refueling interval ([less than or equal

to] 730 days, nominally 24 months). Additionally, the amendment removed

a footnote for TS 4.5.2.b that is no longer applicable.

Date of issuance: February 11, 1997.

Effective date: February 11, 1997, to be implemented not later than

120 days over issuance.

Amendment No.: 214.

Facility Operating License No. NPF-3: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: October 9, 1996.

The Commission's related evaluation of the amendment is contained

in a Safety Evaluation dated February 11, 1997.

No significant hazards consideration comments received: No.

Local Public Document Room location: University of Toledo, William

Carlson Library, Government Documents Collection, 2801 West Bancroft

Avenue, Toledo, Ohio 43606.

Dated at Rockville, Maryland, this 19th day of February 1997.

For the Nuclear Regulatory Commission.

Jack W. Roe,

Director, Division of Reactor Projects--III/IV, Office of Nuclear

Reactor Regulation.

[FR Doc. 97-4573 Filed 2-25-97; 8:45 am]

BILLING CODE 7590-01-P