Applications and Amendments to Facility Operating LicensesInvolving No Significant Hazards Considerations

NUCLEAR REGULATORY COMMISSION

Biweekly Notice

Applications and Amendments to Facility Operating

LicensesInvolving 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 July 19, 1997, through August 1, 1997. The

last biweekly notice was published on July 30, 1997, (62 FR 40843).

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 September 12, 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

[[Page 43366]]

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

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

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

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

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

Nontimely filings of petitions for leave to intervene, amended

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

not be entertained absent a determination by the Commission, the

presiding officer or the Atomic Safety and Licensing Board that the

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

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

For further details with respect to this action, see the

application for amendment which is available for public inspection at

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

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

the particular facility involved.

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: July 8, 1997

Description of amendments request: The proposed amendments remove

the suppression chamber water volume band from Technical Specification

(TS) 3.6.2.1.a.1 while retaining the equivalent water level band. The

values for the suppression chamber water volume corresponding to the

low and high suppression chamber water levels will be retained in the

Bases section of the TS and will be revised by the proposed amendments

to account for the displacement of water due to the planned

installation of larger emergency core cooling system suction strainers.

The revised relationship between the high and low suppression chamber

water levels and suppression chamber water volume will also be

described in the Updated Final Safety Analysis Report.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below: 1. The proposed amendments do

not involve a significant increase in the probability or consequences

of an accident previously evaluated.

The proposed change revises the values of the minimum and

maximum suppression chamber pool water volume limits. The water

inventory of the suppression chamber pool is not a precursor of an

accident and, therefore, cannot increase the probability of an

accident previously evaluated. The pressure suppression chamber

water pool mitigates the consequences of loss-of-coolant accidents

(LOCAs) transients [sic], and other events by providing a heat sink

for reactor primary system energy releases. The proposed minimum and

maximum pool water volume values will be consistent with the current

suppression chamber pool water level limits. No changes to setpoints

will be made as a result of the proposed change. The impact of the

proposed change to the minimum and maximum suppression chamber pool

volume limits on the suppression chamber pool temperatures and

pressures following a design basis LOCA, an Safety/Relief Valve

(SRV) blowdown event, an Anticipated Transient Without Scram (ATWS)

event, an Appendix R fire event, and a station blackout event has

been evaluated and does not cause accident parameters to exceed

acceptable values. In addition, the impact the proposed change has

on the time to reach cold shutdown when using the alternate Residual

Heat Removal (RHR) shutdown cooling function is negligible. The

potential impact the proposed change to the suppression chamber pool

water volume limits has on SRV line loads, SRV discharge line

reflood height, wetwell pressurization, suppression chamber pool

swell loads, vent thrust loads, and condensation oscillation and

chugging loads was also reviewed. The change to the suppression

chamber pool water volume limits has no significant adverse impact

on any of these parameters. As delineated above, the capability of

the suppression chamber water pool to perform its mitigative

functions is not affected by the proposed change. Therefore, the

proposed change does not involve a significant increase in the

consequences of an accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

[[Page 43367]]

The proposed change revises the values of the minimum and

maximum volume of the suppression chamber water pool. The proposed

change will not alter any physical mechanism by which the

suppression chamber water pool volume is maintained between the

minimum and maximum values. The suppression chamber pool water level

will continue to be maintained between -27 and -31 inches. The

suppression chamber pool water level limits are retained in

Technical Specification (TS) 3.6.2.1.a.1, since this is the

information available to the operators regarding the suppression

chamber pool water volume limits. These level limits are equivalent

to the suppression chamber pool water volume limits; therefore, it

is only the presentation of the equivalency that is being relocated

to the Bases and the Updated Final Safety Analysis Report (UFSAR).

As such, the relocated suppression chamber pool water volume limits

are not required to be in the TS to provide adequate protection of

the public health and safety. As a result of the proposed strainer

changes, there are no physical changes to any other suppression

chamber components or instrumentation. No new mode of operation is

introduced as a result of the proposed change. Analyses have been

performed which conclude that the proposed change will not affect

the operability of the equipment designed to mitigate the

consequences of an accident. 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 license amendments do not involve a significant

reduction in a margin of safety.

The proposed change revises the values of the minimum and

maximum suppression chamber water pool volumes. The pressure

suppression chamber water pool mitigates the consequences of several

postulated accidents and transients by providing a heat sink for the

primary coolant system. These accidents and events are the

postulated design basis LOCA, an SRV blowdown event, an ATWS event,

an Appendix R fire, and station blackout events. The consequences of

the change in the suppression pool water volume limits have been

evaluated for these events, and there is no significant reduction in

the margin of safety.

The results of the analyses for the postulated accidents and

events indicate the temperature of the suppression chamber pool

water could increase slightly as a consequence of the decrease in

the minimum suppression chamber pool water volume limit. However,

the suppression chamber pool water and containment temperatures

remain within acceptable values. The impact of the calculated

increase in containment temperature on the available Net Positive

Suction Head (NPSH) for the Residual Heat Removal (RHR) and Core

Spray pumps has been evaluated for the postulated design basis LOCA

and indicate[s] adequate NPSH is maintained throughout the event.

The potential impact of the proposed change to the suppression

chamber pool water volume limits on the SRV line loads, SRV

discharge line reflood height, wetwell pressurization, suppression

chamber pool swell loads, vent thrust loads, and condensation

oscillation and chugging loads was evaluated with the conclusion

that there are no adverse impacts on these parameters.

In addition, a small suppression chamber pool water temperature

increase could result due to the reduction in minimum suppression

pool volume limit in the event reactor shutdown is conducted through

a path utilizing the suppression chamber pool. Such a shutdown path

is an alternative to the normal RHR shutdown cooling function, and

the small potential increase in temperature results in a negligible

increase in the time required to reach cold shutdown conditions.

Cold shutdown conditions can still be reached well within the

Technical Specification requirements.

The proposed increase in the suppression pool water volume limit

does not adversely impact containment parameters as a result of

postulated accidents and events. The potential increase in

temperature of the pressure suppression chamber pool water does not

significantly decrease the ability to maintain containment

parameters within acceptable limits. The potential increase in time

to reach cold shutdown conditions utilizing the suppression pool as

an alternative to the normal RHR shutdown cooling function is

negligible. Therefore, the proposed change to revise the minimum and

maximum suppression water pool volumes does not involve a

significant reduction in a margin of safety.

The suppression chamber pool water level limits are retained in

TS 3.6.2.1.a.1, since this is the information available to the

operators regarding the suppression chamber pool water volume

limits. These level limits are equivalent to the suppression chamber

pool water volume limits and the equivalency is being relocated to

the Bases and the UFSAR. As such, the relocated suppression chamber

pool water volume limits are not required to be in the TS to provide

adequate protection of the public health and safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: University of North Carolina

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

Wilmington, North Carolina 28403-3297.

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

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

Raleigh, North Carolina 27602

NRC Project Director: Gordon E. Edison, Acting

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One,

Unit No. 2, Pope County, Arkansas

Date of amendment request: July 21, 1997

Description of amendment request: Technical Specification Change

Request Concerning Emergency Feedwater Surveillance Testing. This

request is to make several changes to the ANO-2 Technical

Specifications including an extension of the emergency feedwater (EFW)

pump surveillance testing frequency, a reduction in the minimum steam

generator pressure required to perform the surveillance testing on the

turbine-driven EFW pump, and a modification to the EFW pump testing

requirements.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does not Involve a Significant Increase in the Probability or

Consequences of an Accident Previously Evaluated.

The proposed changes included in this amendment request are

being made to the emergency feedwater (EFW) system technical

specification (TS) surveillances. These changes include surveillance

interval modifications, allowances to perform the turbine driven EFW

pump surveillance at a lower steam generator (S/G) pressure,

removing the requirements to perform specific EFW surveillance

requirements (SRs) during plant shutdowns, bases changes, and

various administrative changes. These changes are consistent with

the applicable SRs located in NUREG-1432 and have therefore, been

previously approved by the NRC.

These changes do not alter the functional characteristics of any

plant component and do not allow any new modes of operation of any

component. The accident mitigation features of the plant are not

affected by the proposed amendment request. No modifications have

been made to the EFW system due to this amendment request. Although

the minimum steam generator pressure has been reduced for the

turbine driven EFW pump testing, calculations show that significant

margin exists between the proposed value and that needed to

adequately perform the test. The capability of the EFW pumps to

perform their required safety function is not impacted by this

change. The addition of the electric driven EFW flow path

verification will help [to] assure proper alignment of both trains

of EFW following extended outages.

The accident mitigation features of the plant are not affected

by the proposed amendment. No modification has been made to the pump

or turbine driver. The capability of the turbine driven EFW pump to

perform its required function is not impacted by this change. The

EFW pumps will be tested in accordance with the more restrictive of

the

[[Page 43368]]

data points required by the safety analysis or the inservice testing

program. Therefore, this change does not involve a significant

increase in the probability or consequences of any accident

previously evaluated.

2. Does Not Create the Possibility of a New or Different Kind of

Accident from any Previously Evaluated.

No new possibility for an accident is introduced by modifying

the proposed specifications for the surveillance testing of the EFW

pumps. The EFW surveillance requirements will continue to

demonstrated the pump's ability to perform its safety function. The

modifications to the proposed EFW surveillance requirements are

consistent with the current revision of NRC approved NUREG -1432,

``Standard Technical Specifications Combustion Engineering Plants''

(ITS). Therefore, this change does not create the possibility of a

new or different kind of accident from any previously evaluated.

3. Does Not Involve a Significant Reduction in Margin of Safety.

The safety function of the EFW system is not altered as a result

of this change. The capability of the EFW pumps to perform their

required function is not impacted by this change. The capability of

the EFW pumps is not impacted by this change. The EFW pumps will be

tested and proven operable in accordance with the more restrictive

of the data points required by the safety analysis of the inservice

testing program. The addition of the electric driven EFW flow path

verification will help assure [to] proper alignment of both trains

of EFW following extended outages. Therefore, this change does not

involve a significant reduction in the margin of safety.

Therefore, based upon the reasoning presented above and the

previous discussion of the amendment request, Entergy Operations has

determined that the requested change does not involve significant

hazards consideration.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Tomlinson Library, Arkansas

Tech University, Russellville, AR 72801

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

Strawn, 1400 L Street, N.W., Washington, DC 20005-3502

NRC Project Director: James Clifford, Acting

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

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

Date of amendment request: June 26, 1997

Description of amendment request: The proposed amendment would

revise the Operating License No. DPR-72, License Condition 2.C.(5) and

delete the requirement for installation and testing of flow indicators

in the emergency core cooling system to provide indication of 40

gallons per minute flow for boron dilution from the license. Approval

of this amendment will allow removal of the appropriate flow

indicators, DH-45-Fl and DH-46-Fl, from the Crystal River 3 (CR3) Final

Safety Analysis Report.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

Criterion 1

The change does not involve a significant increase in the

probability or consequences of an accident previously evaluated.

This license amendment removes the requirement for flow

indication on the DH drop line and auxiliary pressurizer spray line

for boron precipitation mitigation during a LOCA [Loss of Coolant

Accident]. The original need for these indicators was to provide

flow indication to the operator to aid in decision making relative

to an alternate active method for boron precipitation prevention.

Alternate active methods have been replaced by the passive flow path

through the gaps which exist between the reactor vessel and the

reactor vessel internals. Since auxiliary pressurizer spray flow is

no longer used, and no other active means is required to be employed

by the operator in the event drop line flow is not indicated, the

original usefulness of and need for this indication no longer

exists. Removal of this requirement from the license condition does

not involve a change in the Improved Technical Specifications. The

operators do not use the flow indication for decision making in

post-accident conditions. Since these instruments are no longer used

for boron precipitation mitigation during a LOCA, abandonment or

removal of flow indicator DH-45-Fl and DH-46-Fl does not increase

the probability of an accident because no previously evaluated

accidents at CR-3 are initiated by DH-45-Fl or DH-46-Fl. Those CR-3

accidents that are analyzed are contained in the Final Safety

Analysis Report (FSAR) and include events such as Loss-of-Coolant

Accidents, Main Steam Line Breaks, Station Blackout, Anticipated

Transients Without Scram, etc. Since DH-45-Fl and DH-46-Fl are

attached to the outside of the DH drop line and auxiliary

pressurizer spray line, their removal will not change the design,

material, or construction standards applicable to the DH System

piping. The removal of the indicator will not affect overall system

performance of the ECCS. All of these previously evaluated accidents

described in the CR-3 FSAR have dose consequences which remain well

within the requirements of 10 CFR Part 100 (25 rem whole body, 300

rem thyroid) and GDC [General Design Criterion] 19 (5 rem whole

body, or its equivalent to any part of the body). Removal of DH-45-

Fl and DH-46-Fl will not alter any assumptions made in evaluating

the radiological consequences of any accident described in the FSAR

nor will it affect any fission product barriers since the ECCS and

containment systems will still perform to meet design requirements.

Therefore, removal of DH-45-Fl and DH-46-Fl will not alter the

consequences of an accident previously evaluated.

Criterion 2

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

kind of accident from any accident previously evaluated.

The proposed license amendment removes the requirement for

indicators which were originally installed to aid the operator in

decision making relative to an alternate flow path for boron

precipitation mitigation during a LOCA. These indicators no longer

serve this purpose, since alternate active flow paths are no longer

considered. Evaluations which consider boron precipitation no longer

rely on three active methods of mitigation, but rather one active

and one passive. Operator action is not required to effect the

backup method in the event that the primary method fails due to a

single active failure. The flow indicators are external to the DH

System piping. They do not penetrate any piping so their removal

cannot create the possibility of a new or different kind of

accident. The accident mitigation strategies remain the same

regardless of whether or not the flow indicators are present.

Therefore, the flow indicators serve no purpose in the analyses. The

proposed amendment does not affect any of the parameters or

conditions that could contribute to the initiation of any accidents.

Criterion 3

The change does not involve a significant reduction in the

margin of safety.

Boron precipitation within the reactor vessel during post-LOCA

conditions, if it were to occur, would challenge the margin of

safety that is provided by assuring compliance with Criterion 5 of

10 CFR 50.46. The license amendment does not change the methodology

of mitigating the consequences of boron precipitation following a

LOCA as described in the current licensing basis. The primary method

of flow through the DH drop line and the use of gap flow as the

``backup'' method for prevention of boron precipitation have been

analyzed, shown to meet all the criteria of 10 CFR 50.46, and

accepted by the NRC. The passive method requires no specific

operator action for initiation, in the event that the primary method

fails due to a single active failure. Therefore, the indication

serves no safety function and does not involve a significant

reduction in the margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 34428

[[Page 43369]]

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

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

33733-4042

NRC Project Director: Frederick J. Hebdon

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

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

Date of amendment request: July 18, 1997

Description of amendment request: The proposed amendment would

revise the Crystal River 3 (CR-3) technical specifications (TS) to

incorporate a new TS 3.4.11 for a Low Temperature Overpressure

Protection (LTOP) System. The proposed changes would be consistent with

the recommendations in the NRC Generic Letter 88-11, ``NRC Position on

Radiation Embrittlement of Reactor Vessel Materials and Its Impact on

Plant Operations.'' TS 3.5.3 and associated TS Bases would also be

revised to reflect the proposed change.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. Does Not Involve a Significant Increase in the Probability or

Consequences of an Accident Previously Evaluated.

This change does not involve a significant increase in the

probability or consequences of any accident previously evaluated.

There are currently no LTOP requirements in the CR-3 Improved

Technical Specifications. CR-3 currently implements LTOP features

through administrative controls and a lowered PORV [power-operated

relief valve] setpoint. The proposed change will establish new LTOP

technical specification requirements necessary to preclude an LTOP

event from occurring. The proposed LTOP requirements are based on

safety analyses that apply ASME [American Society of Mechanical

Engineers] Code Case N-514. These requirements will decrease the

probability of a low temperature overpressure event by providing

protection for all pressure and temperature combinations for which a

low temperature overpressure event may be postulated.

The consequences of a low temperature overpressure accident are

not affected by this change. There is no change to the 10 CFR [Code

of Federal Regulations] Part 100 dose calculation for a low

temperature overpressure accident.

2. Does Not Create the Possibility of a New or Different Kind of

Accident from any Previously Evaluated

This change does not create the possibility of a new or

different kind of accident from any previously evaluated.

The new LTOP Technical Specification does not require

modification to the plant nor does it create a new mode of plant

operation. The LTOP system adds no new accident initiators.

3. Does Not Involve a Significant Reduction in the Margin of

Safety.

The proposed change does not involve a significant reduction in

the margin of safety and will provide added safety benefit gained

through the requirements to preclude a low temperature

overpressurization event to the RCS [reactor coolant system].

The margin of safety prior to having an LTOP system was limited

due to the informal, administrative method of minimizing the impact

of a low temperature overpressure accident. By formalizing these

requirements into a technical specification, at the least, margin of

safety is retained and perhaps improved due to the elevated

significance of required actions.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 34428

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

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

33733-4042

NRC Project Director: Frederick J. Hebdon

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

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

Date of amendment request: July 29, 1997

Description of amendment request: The proposed amendment would

revise the Crystal River Nuclear Generating Unit 3 (CR3) technical

specifications (TS) to add subcooling margin and decay heat removal

(low pressure injection) flow and correct certain nomenclature in the

post-accident monitoring (PAM) instrumentation TS. In addition, the

licensee proposes to add emergency diesel generator (EDG) kilowatt (kW)

indication to the PAM instrumentation. Specifically, the following TS

would be revised:

A. Table 3.3.17-1, Function 8: The descriptor is changed from

``Containment Pressure (Narrow Range)'' to ``Containment Pressure

(Expected Post-Accident Range).''

B. Table 3.3.17-1, Function 18: The required channels for Core Exit

Temperature (Backup) is changed from ``2 sets of 5'' to ``3 per core

quadrant.''

C. Table 3.3.17-1: A new Function 20 is added and designated as

``Low Pressure Injection Flow'', with 2 required channels, and

Condition E.

D. Table 3.3.17-1: A new Function 21 is added and designated as

``Degrees of Subcooling'', with 2 required channels, and Condition E.

E. Table 3.3.17-1: A new Function 22 is added and designated as

``Emergency Diesel Generator kW Indication'', with 2 required channels,

and Condition E. A note clarifying the number of required channels is

added: ``(c): one indicator per EDG''.

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 items A, B, C, D and E

corresponds to the specific TS changes described above.

1. The proposed changes will not significantly increase the

probability or consequences of an accident previously evaluated

because:

A/B. The changes in containment pressure and core exit

thermocouple nomenclature do not reflect any physical changes to the

facility. This would have no impact on accident probability or

consequences.

C/D/E. The addition of low pressure injection flow, degrees of

subcooling, and EDG kW indication to the Post-Accident

Monitoring Instrumentation LCO [Limiting Condition for Operation] is

being done to comply with a commitment made during the technical

specification improvement program to include in the technical

specifications that instrumentation which monitors variables

classified as Type A in accordance with Regulatory Guide 1.97. These

three variables have been reclassified as Type A. The associated

instruments are used in post-accident conditions to prompt the

operators to take certain mitigative actions. Therefore, the

probability of an accident occurring is unaffected. As part of the

re-classification of these variables to Type A and inclusion in

technical specifications, the associated monitoring instrumentation

will be under more strict surveillance and control, which provides

additional assurance that the prescribed manual operator actions

will be implemented when necessary. This, in turn, assures the

previously evaluated accident consequences remain valid.

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

or different kind of accident from any accident previously evaluated

because:

A/B. The changes in containment pressure and core exit

thermocouple nomenclature do not reflect any physical changes to

the facility. The changes provide clarification for the instruments

which are required to comply with the LCO. This would not create

possibility of a new or different kind of accident.

C/D/E.The addition of low pressure injection flow, degrees of

subcooling, and EDG kW indication to the Post-Accident Monitoring

Instrumentation LCO is being

[[Page 43370]]

done to comply with a commitment made during the technical

specification improvement program to include in the technical

specifications that instrumentation which monitors variables

classified as Type A in accordance with Regulatory Guide 1.97. These

three variables have recently been reclassified as Type A. The

associated instruments are used after an accident occurs to prompt

the operators to take certain mitigative actions. Since the

instrumentation is used only post-accident, these changes do not

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

3. The proposed change will not involve a significant reduction

to the margin of safety because:

A/B. The changes in containment pressure and core exit

thermocouple nomenclature have no affect on the margin of safety.

The changes provide clarification of the technical specifications.

This reduces the potential for confusion regarding this

instrumentation.

C/D/E. The addition of low pressure injection flow, degrees of

subcooling, and EDG kW indication to the post-accident

monitoring instrumentation table in technical specifications results

in added controls on the OPERABILITY of this post-accident

monitoring instrumentation and provides greater assurance that it

will be available should an accident occur.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 34428

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

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

33733-4042

NRC Project Director: Frederick J. Hebdon

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: July 18, 1997

Description of amendment request: The proposed amendment adds a new

Technical Specification and associated Bases to address the operability

of the steam generator atmospheric relief bypass valves (SGARBVs).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

NNECO has reviewed the proposed revision in accordance with

10CFR50.92 and has concluded that the revision does not involve a

significant hazards consideration (SHC). The basis for this

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

satisfied. The proposed revision does not involve an SHC because the

revision would not:

1. Involve a significant increase in the probability or

consequence of an accident previously evaluated.

The operability of the SGARBVs provides a method to recover from

a SGTR [steam generator tube rupture] event during which the

operator is required to perform a limited cooldown to establish

adequate subcooling as a necessary step to limit the primary to

secondary break flow into the ruptured steam generator. For other

design events, the SGARBVs provide a safety grade method for cooling

the unit to residual heat removal entry conditions should the

preferred heat sink via the steam bypass system or the steam

generator atmospheric relief valves be unavailable. This proposed

revision to the Technical Specifications will add a new Technical

Specification 3/4.7.1.6 and its associated Bases Section 3/4.7.1.6

which were developed bases on the information contained in the

Westinghouse Improved Standard Technical Specifications, NUREG 1431,

Rev. 1. The proposed specification and bases provide further

assurance that the SGARBVs will be available to function as

described in the accident analysis.

Therefore, the proposed revision 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.

This proposed revision to the Technical Specifications to add a

new specification and bases for the SGARBVs does not cause a change

in the operation of any system or component during normal or

accident conditions.

Therefore, the proposed revision does not create the possibility

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

evaluated.

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

The proposed new Technical Specification 3/4.7.1.6 and its

associated Bases Section 3/4.7.1.6 were developed based on the

information contained in the Westinghouse Improved Standard

Technical Specifications, NUREG 1431, Rev. 1. The SGARBV's are not

currently in the Technical Specifications of Millstone Unit No. 3

and are being added to ensure accident mitigation functional

capability. The NUREG 1431, Rev. 1 surveillance frequency is 18

months. The NUREG 1431, Rev. 1 surveillance frequency bases reads

``operating experience has shown that these components usually pass

the surveillance when performed at the 18 month frequency''. The

proposed frequency acceptability has been evaluated by reviewing

SGARBV AWO's [automated work order's] for the period from Jan. 1990

to April 1997 to confirm the absence of excessive work orders which

indicate valve functional failures and none were identified.

Additionally, each SGARBV line consists of one SGARBV and an

associated block valve. These proposed changes are consistent with

the design and operation of the SGARBVs. There is no negative affect

on the dose consequences from any design basis event or core damage

frequency.

Therefore, the proposed revision does not involve a significant

reduction in a margin of safety.

In conclusion, based on the information provided, it is

determined that the proposed revision does not involve an SHC.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Learning Resources Center,

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

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

49 Rope Ferry Road, Waterford, Connecticut

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

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

CT 06141-0270

NRC Deputy Director: Phillip F. McKee

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

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

County, Minnesota

Date of amendment requests: November 27, 1996

Description of amendment requests: The proposed amendment[s] would

incorporate new steam generator tube sleeve designs and installation

and examination techniques into the Prairie Island Technical

Specifications.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

increase in the probability or consequences of an accident

previously evaluated.

The supporting technical evaluation and safety evaluation for

the Combustion Engineering leak tight sleeves demonstrate that the

sleeve configuration will provide steam generator tube structural

and leakage integrity under normal operating and accident

conditions. The sleeve configurations have been designed and

analyzed in accordance with the requirements of the ASME [American

Society of Mechanical Engineers] Code. Mechanical testing has shown

that the sleeve and sleeve joints provide margin above acceptance

[[Page 43371]]

limits. Ultrasonic examination is used to verify the leak tightness

of the above the [sic] tubesheet sleeve welds. Testing has

demonstrated the leak tightness of the hard roll joint as well as

the structural integrity of the hard roll joint. Tube rupture can

not occur at the hard roll joint due to the reinforcing effect of

the tubesheet. Tests have demonstrated that tube collapse will not

occur due to postulated LOCA [loss-of-coolant accident] loadings.

The existing Technical Specification leakage rate requirements

and accident analysis assumptions remain unchanged in the event that

significant leakage did occur from the sleeve joints or that a

sleeve assembly ruptured. Any leakage through the sleeve assembly is

fully bounded by the existing steam generator tube rupture analysis

included in the Prairie Island Plant USAR [updated safety analysis

report]. The proposed sleeving repair does not adversely impact any

other previously evaluated design basis accident.

The sleeve minimum acceptable wall thickness used for developing

the depth based plugging limit for the sleeve is determined using

the guidance of draft Regulatory Guide 1.121 [Bases for

Plugging Degraded PWR [Pressurized-Water Reactor] Steam Generator

Tubes] and the pressure stress equation of Section III of

the ASME Code. Evaluation of the minimum acceptable wall thickness

for normal, upset, and postulated accident condition loading per the

ASME Code finds that the limiting condition is established from

normal operating conditions which then bounds the upset and accident

condition values. Allowance for non-destructive examination and

growth of existing sleeve wall degradation must be made when

determining the sleeve plugging limit. The proposed plugging limit

is 40% through wall degradation. The sleeve assembly will be

examined by state of the art non-destructive examination techniques

on a periodic basis to provide early indication of sleeve

degradation. The corrosion resistance of the Alloy 690 sleeve has

been verified by field experience at Prairie Island. The oldest

Alloy 690 sleeves were installed May 1987. No indication of

corrosion of the sleeve or the parent tube in the weld joint has

been identified by state-of-the-art eddy current examination. These

oldest sleeve welds did not receive post weld heat treatment. In

addition, 5 sleeves were removed for destructive examination in

February, 1996. No corrosion was found in any of these sleeves

including those dating from October 1992. The pulled sleeves had

received post weld heat treatment. Post weld heat treatment can be

optionally applied to the free span sleeve weld joints to reduce the

susceptibility of the weld joint and parent tube to stress corrosion

cracking. Since the sleeve design meets the requirements of the ASME

code and mechanical tests have demonstrated margins above acceptance

criteria, the installation of the Combustion Engineering leak tight

sleeves will not increase the probability or consequences of an

accident previously evaluated.

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

a new or different kind of accident from any accident previously

analyzed.

Installation of sleeves does not introduce any significant

changes to the plant design basis. The use of a sleeve to span a

degraded region of steam generator tubing restores the structural

and leakage integrity of the tubing to meet the original design

bases. Stress and fatigue analysis of the sleeve assembly shows that

the requirements for ASME Code are met. Mechanical testing has

demonstrated that margin exists above the design criteria. Any

hypothetical accident as a result of any degradation in the sleeved

tube would be bounded by the existing tube rupture accident

analysis.

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

reduction in the margin of safety.

The use of the sleeves to repair degraded steam generator tubing

has been demonstrated to maintain the integrity of the tube bundle

commensurate with the requirements of the ASME Code and draft

Regulatory Guide 1.121 and to maintain the primary to secondary

pressure boundary under normal and postulated accident conditions.

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

sleeve assembly are consistent with the safety factors in the ASME

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

operational and faulted condition stresses and cumulative fatigue

usage are bounded by the ASME Code requirements. The sleeve assembly

has been verified by testing to prevent both tube pullout and

significant leakage during normal and postulated accident

conditions. A test program was conducted to ensure the rolled joint

design for the lower joint in the tubesheet sleeve was leak tight

and capable of withstanding the designs loads. The primary coolant

pressure boundary of the sleeve assembly will be periodically

inspected by non-destructive examination to identify sleeve

degradation due to operation. Installation of sleeves will decrease

the number of tubes which must be taken out of service. There is a

small amount of primary coolant flow reduction due to sleeves for

which an equivalent plugging sleeve to plug ratio is assigned and is

used to assess the final equivalent plugging percentage used as an

input to other safety analyses. Because the sleeve maintains the

design basis requirements for the steam generator tubing, it is

concluded that the proposed change does not result in a significant

reduction in margin with respect to plant safety as defined in the

USAR or the Technical Specification Bases.

Based on the evaluation described above, and pursuant to 10 CFR

Part 50, Section 50.91, Northern States Power Company has determined

that operation of the Prairie Island Nuclear Generating Plant in

accordance with the proposed license amendment request does not

involve any significant hazards considerations as defined by NRC

regulations in 10 CFR Part 50, Section 50.92.

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

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

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

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401

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

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

NRC Project Director: John N. Hannon

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

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

County, Minnesota

Date of amendment requests: May 15, 1997

Description of amendment requests: The proposed amendments would

change the Technical Specifications (TS) to revise certain limitations

on reactor coolant system leakage and steam generator tube

surveillance. The proposed changes would implement a voltage-based

repair criteria per the requirements of NRC Generic Letter 95-05,

``Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes

Affected by Outside Diameter Stress Corrosion Cracking.'' In addition,

a typographical error in TS Section 4.12.c. is being corrected.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

increase in the probability or consequences of an accident

previously evaluated.

The supporting technical evaluation and safety evaluation for

the voltage based repair criteria demonstrate that steam generator

tube structural and leakage integrity under normal operating and

accident conditions will be maintained. Tube burst criteria are

inherently satisfied during normal operating conditions due to the

proximity of the tube support plate (TSP). Test data referenced in

Generic Letter 95-05 indicates that tube burst cannot occur within

the TSP, even for tubes which have 100% throughwall electric

discharge machining notches, 0.75 inch long, provided that the TSP

is adjacent to the notched area. Since tube-to-TSP proximity

precludes tube burst during normal operating conditions, use of the

criteria must retain tube integrity characteristics which maintain a

margin of safety of 1.43 times the bounding faulted condition, main

steamline break (MSLB) pressure differential. The Regulatory Guide

(RG) 1.121 [Bases for Plugging Degraded PWR [Pressurized-

Water Reactor] Steam Generator Tubes] criterion requiring

maintenance of a safety factor of 1.43 times the MSLB pressure

differential on tube burst

[[Page 43372]]

is satisfied by 7/8'' diameter tubing with bobbin coil indications

with signal amplitudes less than the current 8.7 volts structural

limit, regardless of the indicated depth measurement.

The upper voltage repair limit (VURL) will be

determined prior to each outage using the most recently NRC approved

database to determine the tube structural limit (VSL).

The structural limit is reduced by allowances for nondestructive

examination (NDE) uncertainty (VNDE) and growth

(VGR) to establish VURL. Using the Generic

Letter (GL) 95-05 NDE and growth allowances for an example, the NDE

uncertainty component of 20% and a voltage growth allowance of 30%

per full power year can be utilized to establish a VURL

of 5.2 volts.

Relative to the expected leakage during accident condition

loadings, it has been previously established that a postulated MSLB

outside of containment but upstream of the main steam isolation

valve (MSIV) represents the most limiting radiological conditions to

the plugging criteria. In support of [the] implementation of the

revised plugging limit, analyses will be performed to determine

whether the distribution of cracking indications at the tube support

plate intersections during future cycles are projected to be such

that primary-to secondary leakage would result in postulated off

site and control room doses exceeding the limits established for

application of the voltage-based repair criteria at Prairie Island.

A separate calculation has determined the maximum allowable MSLB

leakage limit in a faulted loop. This limit was calculated using the

technical specification reactor coolant system (RCS) Iodine-131

activity level of 1.0 microcuries per gram dose equivalent Iodine-

131 and the recommended Iodine-131 transient spiking values

consistent with NUREG-0800 [Standard Review

Plan]. The projected MSLB leak rate calculation

methodology prescribed in Section 2.b of Generic Letter 95-05 will

be used to calculate the end-of-cycle (EOC) leakage. Projected EOC

voltage distribution will be developed using the most recent EOC

eddy current results and considering an appropriate voltage

measurement uncertainty and indication growth allowance. The log-

logistic probability of leakage correlation will be used to

establish the MSLB leak rate used for comparison with the faulted

loop allowable limit. Therefore, as implementation of the voltage-

based repair criteria does not adversely affect steam generator tube

integrity and implementation will be shown to result in acceptable

dose consequences, the proposed amendment[s] [do] not result in any

increase in the probability or consequences of an accident

previously evaluated in the Updated Safety Analysis Report (USAR).

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

a new or different kind of accident from any accident previously

analyzed.

Implementation of the proposed steam generator tube voltage-

based repair criteria does not introduce any significant changes to

the plant design basis. Use of the voltage-based repair criteria

does not provide a mechanism which could result in an accident

outside of the region of the tube support plate elevations since

tubes with outside diameter stress corrosion cracking (ODSCC) not

occurring inside the thickness of the tube support plates will be

plugged or repaired. Neither a single or multiple tube rupture event

would be expected during all plant conditions in a steam generator

in which the voltage based repair limit has been applied.

Northern States Power will implement a maximum primary-to-

secondary leak rate limit of 150 gpd [gallons per day] per steam

generator to help preclude the potential for excessive leakage

during all plant conditions. The Regulatory Guide 1.121 criterion

for establishing operational leak rate limits that require plant

shutdown are based upon leak-before-break considerations to detect a

free span crack before potential tube rupture during faulted plant

conditions. The 150 gpd limit provides for leakage detection and

plant shutdown in the event of the occurrence of an unexpected

single crack resulting in leakage that is associated with the

longest permissible crack length.

The operational leakage limit will be reduced to 150 gpd limit

consistent with Generic Letter 95-05. This limit is expected to

provide for plant shutdown prior to reaching critical lengths for

MSLB conditions using the lower 95% leak rate data. Additionally,

this leak-before-break evaluation assumes that the entire crevice

area is uncovered during blowdown. Partial uncover will provide

benefit to the burst capacity of the intersection and only a small

percentage of the TSPs are deflected greater than the TSP thickness

during a postulated MSLB.

As steam generator tube integrity upon implementation of the

voltage-based repair criteria continues to be maintained through

inservice inspection and primary-to secondary leakage monitoring,

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

accident previously evaluated is not created.

Therefore, the proposed change does not create the possibility

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

evaluated.

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

reduction in the margin of safety.

The use of the voltage-based repair criteria at Prairie Island

maintains steam generator tube integrity commensurate with the

criteria of the ASME [American Society of Mechanical Engineers] Code

and Regulatory Guide 1.121. Regulatory Guide 1.121 describes a

method acceptable to the Commission for meeting GDCs [General Design

Criteria] 14, 15, 30, 31, and 32 by reducing the probability or the

consequences of steam generator tube rupture. This is accomplished

by determining the limiting conditions of degradation of steam

generator tubing, as established by inservice inspection, for which

tubes with unacceptable cracking should be repaired or removed from

service. Upon implementation of the proposed criteria, even under

the worst case conditions, the occurrence of ODSCC at the tube

support plate elevations is not expected to lead to the steam

generator tube rupture event during normal or faulted plant

conditions. The EOC distribution of crack indications at the tube

support plate elevations will be confirmed to result in acceptable

primary-to-secondary leakage during all plant conditions in order to

assure that radiological consequences meet the requirements of

Generic Letter 95-05.

Previous evaluations have indicated a potential for tube

deformation and collapse during a postulated loss-of-coolant

accident (LOCA) plus safe-shutdown-earthquake (SSE) event. The tube

collapse potential arises from TSP deformation at the support plate

wedges. Evaluation of the Westinghouse umbrella seismic spectra

provided in Westinghouse letter NSP-92-152 for Model 51 steam

generators shows that Prairie Island is bounded by those spectra and

that no tubes will undergo deformation due to the combined effects

of LOCA plus SSE. Therefore, no tubes need to be excluded from

application of the voltage based criteria due to deformation

resulting from combined LOCA plus SSE loadings. Addressing

Regulatory Guide 1.83 [Inservice Inspection of

Pressurized Water Reactor Steam Generator Tubes]

considerations, implementation of the voltage-based repair criteria

is supplemented by enhanced eddy current inspection guidelines to

provide consistency in voltage normalization, by an extensive bobbin

coil inspection which will include 100% of the hot leg TSP

intersections and cold leg intersections down to the lowest cold leg

TSP with known ODSCC, by the determination of the TSPs having ODSCC

using at least 20% random sampling of tubes inspected over their

full length, and by rotating pancake coil inspection (or equivalent)

requirements for the larger indications left in service to

characterize the principal degradation as ODSCC.

As noted previously, implementation of the tube support plate

intersection voltage-based repair criteria will decrease the number

of tubes which must be repaired. The installation of steam generator

tube plugs or sleeves reduces the RCS flow margin. Thus,

implementation of the voltage-based repair criteria will maintain

the margin of flow that would otherwise be reduced in the event of

increased tube plugging.

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

amendment request does not result in a significant reduction in

margin with respect to plant safety as defined in the USAR or any

Bases of the plant Technical Specifications.

Based on the evaluation described above, and pursuant to 10 CFR

Part 50, Section 50.91, Northern States Power Company has determined

that operation of the Prairie Island Nuclear Generating Plant in

accordance with the proposed license amendment request does not

involve any significant hazards considerations as defined by NRC

regulations in 10 CFR Part 50, Section 50.92.

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. In addition, the proposed correction to a typographical

error has no effect on the three standards of 10

[[Page 43373]]

CFR 50.92(c). Therefore, the NRC staff proposes to determine that the

amendment requests involve no significant hazards consideration.

Local Public Document Room location: Minneapolis Public Library,

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401

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

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

NRC Project Director: John N. Hannon

PECO Energy Company, Public Service Electric and Gas Company,

Delmarva Power and Light Company, and Atlantic City Electric

Company, Dockets Nos. 50-277 and 50-278, Peach Bottom Atomic Power

Station, Units Nos. 2 and 3, York County, Pennsylvania

Date of application for amendments: June 4, 1997

Description of amendment request: The proposed Technical

Specifications (TSs) amendment revises TS Surveillance Requirement

3.8.2.1 to no longer require that automatic emergency diesel generator

(EDG) auto-start and trip bypass features must be functional when the

emergency core cooling system (ECCS) is not required to be operable.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed change to the facility does not involve a

significant increase in the probability or consequences of an

accident previously evaluated.

The proposed change will eliminate an inconsistency between

Technical Specifications 3.3.5.1, 3.5.2, and 3.8.2 by clarifying

that the EDG auto-start and EDG trip bypass on ECCS initiation

capability is not required during periods in which ECCS is not

required to be OPERABLE. No physical changes to the facility will be

made per this change. The systems, structures, and components

affected by this change are considered to be accident mitigators and

not accident initiators. The affected systems, structures, and

components will continue to operate within the current design

parameters. The ability of the EDGs to auto-start on a loss of

offsite power or degraded voltage will remain unchanged. No new

failure modes or conditions adverse to safety will be created as a

result of this change. Therefore, the proposed change does not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

2. The proposed change to the facility does not create the

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

previously evaluated.

The proposed change will eliminate an inconsistency between

Technical Specifications 3.3.5.1, 3.5.2, and 3.8.2 by clarifying

that the EDG auto-start and EDG trip bypass on ECCS initiation

capability is not required during periods in which ECCS is not

required to be OPERABLE. No physical changes to the facility will be

made per this change. The systems, structures and components

affected are considered to be accident mitigators not accident

initiators. The affected systems, structures and components will

continue to operate within the current design parameters. No new

failure modes or conditions adverse to safety will be created as a

result of this change. The plant conditions which do not require any

ECCS to be OPERABLE, (i.e., the plant in MODE 5, the spent fuel

storage pool gates are removed, water level is greater than or equal

to 458 inches above reactor pressure vessel instrument zero, and

there are no OPDRVs [operations with the potential of draining the

reactor vessel] in progress) ensure sufficient coolant inventory to

allow operator action to prevent uncovering the fuel. The ability of

the EDGs to auto-start on a loss of offsite power or degraded

voltage will remain unchanged. Therefore, the proposed change does

not involve a significant increase in the probability or

consequences of an accident previously evaluated.

3. The proposed change to the facility does not involve a

significant reduction in a margin of safety.

The proposed change will eliminate an inconsistency between

Technical Specifications 3.3.5.1, 3.5.2, and 3.8.2 by clarifying

that the EDG auto-start and EDG trip bypass on ECCS initiation

capability is not required during periods in which ECCS is not

required to be OPERABLE. The ECCS and EDGs capability to perform the

required safety functions as described/required in the bases of the

current plant Technical Specifications will be maintained.

Therefore, the proposed change to the facility does not result in a

significant reduction in any margin of safety.

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

this review, it appears that the three standards of 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 Publications

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

Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,

PA 17105

Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and

General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,

PA 19101

NRC Project Director: John F. Stolz

PECO Energy Company, Public Service Electric and Gas Company,

Delmarva Power and Light Company, and Atlantic City Electric

Company, Docket No. 50-278, Peach Bottom Atomic Power Station, Unit

No. 3, YorkCounty, Pennsylvania

Date of application for amendment: June 30, 1997

Description of amendment request: The proposed amendment would

revise the Technical Specification (TS) 2.1.1.2 safety limit minimum

critical power ratios (SLMCPRs) to be consistent with the use of GE 13

fuel in the Unit 3 core for operating cycle 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 TS changes do not involve a significant increase

in the probability or consequences of an accident previously

evaluated.

The derivation of the cycle-specific SLMCPRs for incorporation

into the TS, and its use to determine cycle-specific thermal limits,

have been performed using the methodology discussed in ``General

Electric Standard Application for Reactor Fuel,'' NEDE-24011-P-A-13,

and U.S. Supplement, NEDE-24011-P-A-13-US, August, 1996, and the

``Proposed Amendment 25 to GE Licensing Topical Report NEDE-24011-P-

A (GESTAR II) on Cycle Specific Safety Limit MCPR.'' Amendment 25

was submitted by GENE to the U.S. Nuclear Regulatory Commission

(USNRC) on December 13, 1996. This change in SLMCPRs cannot increase

the probability or severity of an accident.

The basis of the SLMCPR calculation is to ensure that greater

than 99.9% of all fuel rods in the core avoid transition boiling if

the limit is not violated. The new SLMCPRs preserve the existing

margin to transition boiling and fuel damage in the event of a

postulated accident. The fuel licensing acceptance criteria for the

SLMCPR calculation apply to PBAPS, Unit 3, Cycle 12 in the same

manner as they have applied previously. The probability of fuel

damage is not increased. Therefore, the proposed TS changes do not

involve an increase in the probability or consequences of an

accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

The SLMCPR is a TS numerical value, designed to ensure that

transition boiling does not occur in 99.9% of all fuel rods in the

core during the limiting postulated accident. It cannot create the

possibility of any new type of accident. The new SLMCPRs are

calculated using methodology discussed in ``General Electric

Standard Application for Reactor Fuel,'' NEDE-24011-P-A-13, and U.S.

Supplement, NEDE-24011-P-A-13-US, August, 1996, and the ``Proposed

[[Page 43374]]

Amendment 25 to GE Licensing Topical Report NEDE-24011-P-A (GESTAR

II) on Cycle Specific Safety Limit MCPR.'' Amendment 25 was

submitted by GENE to the U.S. Nuclear Regulatory Commission (USNRC)

on December 13, 1996.

Therefore, the proposed TS changes do not create the possibility

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

evaluated.

3. The proposed TS changes do not involve a significant

reduction in a margin of safety.

The margin of safety as defined in the TS Bases will remain the

same. The new SLMCPRs are calculated using methodology discussed in

``General Electric Standard Application for Reactor Fuel,'' NEDE-

24011-P-A-13, and U.S. Supplement, NEDE-24011-P-A-13-US, August,

1996, and the ``Proposed Amendment 25 to GE Licensing Topical Report

NEDE-24011-P-A (GESTAR II) on Cycle Specific Safety Limit MCPR.''

Amendment 25 was submitted by GENE to the U.S. Nuclear Regulatory

Commission (USNRC) on December 13, 1996. The fuel licensing

acceptance criteria for the calculation of the SLMCPR apply to PBAPS

[Peach Bottom Atomic Power Station], Unit 3 Cycle 12 in the same

manner as they have applied previously. The SLMCPRs ensure that

greater than 99.9% of all fuel rods in the core will avoid

transition boiling if the limit is not violated, thereby preserving

the fuel cladding integrity. Therefore, the proposed TS changes do

not involve a reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Government Publications

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

Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,

PA 17105

Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and

General Counsel, PECO Energy Company, 2301 Market Street, Philadelphia,

PA 19101

NRC Project Director: John F. Stolz

Power Authority of the State of New York, Docket No. 50-333, James

A. FitzPatrick Nuclear Power Plant, Oswego County, New York

Date of amendment request: April 14, 1997

Description of amendment request: The proposed amendment revises

Appendix A, Section 6 of the Technical Specifications. The changes will

enable Safety Review Committee (SRC) to review plant staff performance

by deleting the plant staff performance requirement from Section

6.5.2.9.b and incorporating a plant staff review requirement in Section

6.5.2.8. The amendment also replaces the position title of Vice

President (VP) Regulatory Affairs and Special Projects (RASP) with

Director of Regulatory Affairs and Special Projects.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

(1) Does the proposed license amendment involve a significant

increase in the probability or consequences of an accident

previously analyzed?

Response:

This amendment application does not involve a significant

increase in the probability or consequences of an accident

previously analyzed. The proposed changes allow the SRC to perform a

review, rather than an audit, of plant staff performance. This

change does not diminish the SRC's effectiveness. A review of the

1995 QA [quality assurance] audit of plant staff performance shows

that no findings related to plant staff performance were issued.

This indicates that the other review mechanisms currently in place

are sufficient to ensure that plant staff performance is monitored.

The position title change of VP-RASP to Director-RASP is an

administrative change as all previously performed functions are

being maintained. Therefore, the proposed changes do not affect the

probability or consequences of any previously analyzed accident.

(2) Does the proposed license amendment create the possibility

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

evaluated?

Response:

This amendment application does not create the possibility of a

new or different kind of accident from any accident previously

evaluated. The proposed changes affect an SRC audit requirement and

a management position title. These changes do not affect plant

equipment or the way the plant operates. Therefore, they cannot

create a new or different kind of accident.

(3) Does the proposed amendment involve a significant reduction

in a margin of safety?

Response:

This amendment application does not involve a significant

reduction in a margin of safety. The requested Technical

Specification revisions require the SRC to review rather than audit

facility staff performance and will not diminish the effectiveness

of the SRC. A review of the 1995 audit confirms that performance of

the annual audit is redundant as no findings or recommendations

concerning plant staff performance were made. The QA/ORG quarterly

trend reports and SRC review of facility staff performance are

adequate to ensure that plant staff performance is properly

monitored.

The position title change (VP-RASP to Director-RASP) is an

administrative change as all previously performed functions are

being maintained. Therefore, the proposed changes do not involve a

significant reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Reference and Documents

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

York 13126.

Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New

York, New York 10019

NRC Project Director: Alexander W. Dromerick, Acting Project

Director

Public Service Electric & Gas Company, Docket No. 50-354, Hope

Creek Generating Station, Salem County, New Jersey

Date of amendment request: March 31, 1997, as supplemented by

letter dated July 16, 1997. The July 16, 1997, supplement supersedes

the March 31, 1997 application.

Description of amendment request: The proposed amendment would

provide changes to Technical Specification (TS) 2.1.2, ``THERMAL POWER,

High Pressure and High Flow,'' ACTION a.1.c for TS 3.4.1.1,

``Recirculation Loops,'' and the Bases for TS 2.1, ``Safety Limits.''

These changes are being made to implement an appropriately conservative

Safety Limit Minimum Critical Power Ratio, to include Cycle 8 specific

analyses, for all Hope Creek core and fuel designs.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of an accident previously evaluated.

The derivation of the revised SLMCPRs for Hope Creek for

incorporation into the Technical Specifications, and its use to

determine cyclespecific thermal limits, have been performed using

NRC approved methods. Additionally, interim implementing procedures

which incorporate cyclespecific parameters have been used which

result in a more restrictive value for SLMCPR. These calculations do

not change the method of operating the plant and have no effect on

the probability of an accident initiating event or transient.

There are no significant increases in the consequences of an

accident previously evaluated. The basis of the MCPR Safety Limit is

to ensure that no mechanistic fuel damage is calculated to occur if

the limit is not violated. The new SLMCPRs preserve the

[[Page 43375]]

existing margin to transition boiling and the probability of fuel

damage is not increased. Therefore, the proposed change does not

involve an increase in the probability or consequences of an

accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

The proposed changes contained in this submittal result from an

analysis of the Cycle 7 and Cycle 8 core reloads using the same fuel

types as previous cycles. These changes do not involve any new

method for operating the facility and do not involve any facility

modifications. No new initiating events or transients result from

these changes. Therefore, the proposed Technical Specification

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

accident, from any accident previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The margin of safety as defined in the Technical Specification

bases will remain the same. The new SLMCPRs are calculated using NRC

approved methods which are in accordance with the current fuel

design and licensing criteria. Additionally, interim implementing

procedures, which incorporate cyclespecific parameters, have been

used. The MCPR Safety Limit remains high enough to ensure that

greater than 99.9% of all fuel rods in the core will avoid

transition boiling if the limit is not violated, thereby preserving

the fuel cladding integrity. Therefore, the proposed Technical

Specification changes do not involve a reduction in a margin of

safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070

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

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

NRC Project Director: John F. Stolz

Public Service Electric & Gas Company, Docket No. 50-354, Hope

Creek Generating Station, Salem County, New Jersey

Date of amendment request: April 1, 1997, as supplemented by letter

dated May 30, 1997

Description of amendment request: The proposed amendment would

provide changes to Technical Specifications (TSs) 4.6.1.1, ``Primary

Containment Integrity,'' 3/4.6.1.2, ``Primary Containment Leakage,'' 3/

4.6.1.3, ``Primary Containment Air Locks,'' 4.6.1.5.1, ``Primary

Containment Structural Integrity,'' and 4.6.1.8.2, ``Drywell and

Suppression Chamber Purge System.'' The amendment would also change the

Bases for 3/4.6.1.2, ``Primary Containment Leakage,'' 3/4.6.1.3,

``Primary Containment Air Locks,'' 3.4.6.1.5, ``Primary Containment

Structural Integrity,'' Section 6, ``Administrative Controls,'' and

License Condition 2.D of Facility Operating License NPF-57. A new TS,

6.8.4.e, ``Primary Containment Leakage Rate Testing Program,'' would be

added. These changes modify the TSs and the Facility Operating License

to adopt the performance based containment leak rate testing

requirements (Option B) of 10 CFR Part 50, Appendix J.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of an accident previously evaluated.

Containment leak rate testing is not an initiator of any

accident. The proposed changes do not make any physical changes to

the containment and do not affect reactor operations or the accident

analyses. Therefore, the proposed changes do not involve a

significant increase in the probability of any previously evaluated

accident.

Since the allowable leakage rate is not being changed and since

the analysis documented in NUREG-1493, ``Performance-Based

Containment Leak-Test Program'' concludes that the impact on public

health and safety due to extended intervals is negligible, the

proposed changes will not involve a significant increase in the

consequences of any previously evaluated accident.

Therefore, adoption of a performance-based leakage testing

requirements will provide an equivalent level of safety and does not

involve a significant increase in the probability or consequences of

an accident previously evaluated.

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

or different kind of accident from any accident previously

evaluated.

No physical changes are being made to the plant, nor are there

any changes being made to the operation of the plant as a result of

the proposed changes. In addition, no new failure modes of plant

equipment previously evaluated are being introduced.

Therefore, the proposed amendment will not create the

possibility of a new or different kind of accident from any

previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed changes are based on NRC-accepted provisions and

maintain adequate levels of reliability of containment integrity.

The performance-based approach to leakage rate testing recognizes

that historically good results of containment testing provide

appropriate assurance of future containment integrity. This supports

the conclusion that the impact on the health and safety of the

public as a result of extended test intervals is negligible. Since

the analysis documented in NUREG-1493 confirms that the performance

based schedule continues to maintain a minimal impact on public

risk, it can be concluded that the margin of safety is not

significantly affected by the proposed changes.

Therefore, the proposed amendment will not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey 08070

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

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

NRC Project Director: John F. Stolz

Public Service Electric & Gas Company, Docket No. 50-354, Hope

Creek Generating Station, Salem County, New Jersey

Date of amendment request: July 3, 1997

Description of amendment request: The proposed amendment would

change Technical Specification Table 3.6.3-1, ``Primary Containment

Isolation Valves'' to add valves to the list, therein.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of an accident previously evaluated.

The accidents previously evaluated in the UFSAR [Updated Final

Safety Analysis Report] that could be possibly affected by this

proposal are those involving loss of coolant scenarios such as a

piping or instrument line break. The proposed relief valves,

associated piping and the affected portions of containment

penetration piping are not initiators of those accidents evaluated

in the UFSAR. The proposed relief valves limit the post-accident

maximum expected pressures of the affected piping segments within

ASME [American Society of Mechanical Engineers] code allowables and

system design pressures. The modification does not cause any system

or component to be operated outside of their design rating

[[Page 43376]]

allowed by applicable codes. The proposed relief valves will be

safety-related and Seismic Category I components (except for the

relief valve discharge piping, which will be non-safety related and

seismically analyzed, and will meet the design, material and

construction standards applicable to the affected piping

segments[)].

The proposed modifications do not jeopardize the capability of

the containment isolation valves in the affected penetrations to

close on the receipt of a containment isolation signal or to

mitigate the consequences of design basis accidents evaluated in the

UFSAR. Although the modifications will result in system pressures to

be above their currently established design values, the new peak

operating pressures of the affected piping segments will be limited

to within the requirements of the ASME code. The modification will

not alter any assumptions previously made or change, degrade, or

prevent actions described in or assumed in evaluating the

radiological consequences of the postulated design basis accidents.

Containment structure temperature and pressure limits will not be

exceeded with this modification and the offsite dose consequences

will not be affected.

Therefore these changes will not significantly increase the

probability of an accident previously evaluated, nor involve a

significant increase in the consequences of an accident previously

evaluated.

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

or different kind of accident from any accident previously

evaluated.

Accidents or malfunctions of equipment important to safety

previously evaluated in the UFSAR relating to the proposed

modification involve the single active failure of a containment

isolation valve to close upon receipt of a containment isolation

signal or its failure to limit the containment bypass leakage

following its closure. The proposed modification: 1) does not impact

the automatic closure times of the containment isolation valves; 2)

does not impact their capability to maintain leak tightness during a

postulated design basis accident; and 3) does not adversely impact

the manner in which any system is operated. The proposed

modification does not compromise the UFSAR accident analysis

assumptions and/or limits. The licensing basis safety analysis

limits for all systems important to safety continue to be met.

Furthermore, there is no change in plant testing proposed in this

change request which could initiate an event. Therefore, these

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

of accident from any accident previously evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

The proposed modifications and Technical Specification changes

do not change the design limits, acceptance criteria or accident

analysis assumptions pertaining to the containment isolation valves,

their associated piping or any other safety-related systems,

structures or components. The proposed modification does not impact

the automatic closure times of the containment isolation valves, nor

does it impact their capability to maintain leak tightness during a

postulated design basis accident. For the systems affected by these

penetration modifications, there is no change in system function or

structural integrity introduced with these proposed changes.

Therefore, the changes contained in this request do not result in a

significant reduction in a margin of safety for the containment

isolation capability of Hope Creek.

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: Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070

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

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

NRC Project Director: John F. Stolz

Public Service Electric & Gas Company, Docket No. 50-354, Hope

Creek Generating Station, Salem County, New Jersey

Date of amendment request: July 7, 1997

Description of amendment request: The proposed amendment would

change Technical Specification (TS) 3/4.8.4.2, ``Motor Operated Valves

- Thermal Overload Protection (BYPASSED),'' to relocate the list of

applicable valves (TS Table 3.8.4.2-1) to the Hope Creek (HC)

Generating Station Updated Final Safety Analysis Report (UFSAR).

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

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

the probability or consequences of an accident previously evaluated.

The proposed TS revisions involve: 1) no hardware changes; 2) no

changes to the operation of any systems or components in normal or

accident operating conditions; and 3) no changes to existing

structures, systems or components. The relocation of Technical

Specification Table 3.8.4.2-1 to the UFSAR and existing surveillance

procedures will continue to ensure that safety-related motor-

operated valves (MOVs) are capable of performing their intended

safety functions. Therefore these changes will not significantly

increase the probability of an accident previously evaluated. To the

extent practicable, these proposed changes were developed consistent

with the changes approved by the NRC when developing NUREG-1433,

``Standard Technical Specifications, General Electric Plants, BWR/

4'', with the intent of having this relocated information controlled

in other plant documents subject to 10CFR50.59 provisions. Since the

plant systems associated with these proposed changes will still be

capable of: 1) meeting all applicable design basis requirements; and

2) retain the capability to mitigate the consequences of accidents

described in the HC UFSAR, the proposed changes were determined to

be justified. Therefore, these changes will not involve a

significant increase in the consequences of an accident previously

evaluated.

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

or different kind of accident from any accident previously

evaluated.

Relocation of Technical Specification Table 3.8.4.2-1 to the

UFSAR will not adversely impact the operation of any safety related

component or equipment. Since the proposed changes involve: 1) no

hardware changes; 2) no changes to the operation of any systems or

components; and 3) no changes to existing structures, systems or

components, there can be no impact on the occurrence of any

accident. To the extent practicable, these proposed changes were

developed consistent with the changes approved by the NRC when

developing NUREG-1433, ``Standard Technical Specifications, General

Electric Plants, BWR/4'', with the intent of having this relocated

information controlled in other plant documents subject to

10CFR50.59 provisions. Furthermore, there is no change in plant

testing proposed in this change request which could initiate an

event. Therefore, these changes will not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

3. The proposed change does not involve a significant reduction

in a margin of safety.

Relocation of Technical Specification Table 3.8.4.2-1 to the

UFSAR is consistent, to the extent practicable, with the changes

approved by the NRC when developing NUREG-1433, ``Standard Technical

Specifications, General Electric Plants, BWR/4''. The MOV thermal

overload protection table will reside in the UFSAR and will ensure

that the associated MOVs will be capable of performing their

intended safety functions. Any changes to this UFSAR table will be

subject to the provisions of 10CFR50.59 and a separate safety

evaluation would be developed to support any proposed changes that

would subsequently be made. Therefore, the changes contained in this

request do 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: Pennsville Public Library, 190

S. Broadway, Pennsville, NJ 08070

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

- N21,

[[Page 43377]]

P.O. Box 236, Hancocks Bridge, NJ 08038

NRC Project Director: John F. Stolz

Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296,

Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County,

Alabama

Date of amendment request: June 2, 1997 (TS 387)

Description of amendment request: The proposed amendment allows

continued plant operation with a single reactor recirculation loop in

service. The Nuclear Regulatory Commission has previously determined

single loop operation is generically acceptable as set forth in Generic

Letter 86-09, ``Technical Resolution of Generic Issue B-59-(N-1) Loop

Operation in BWRs [boiling water reactors] and PWRs [pressurized-water

reactors].'' Single loop operation is also recognized as a standard

mode of operation in the BWR/4 Improved Standard TS.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration, which is presented below:

1. The proposed amendment does not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

An analysis of the limiting operational transients has been

performed by GE [General Electric] for BFN as documented in NEDO-

24236 to demonstrate adequate margin to the Safety Limit Minimum

Critical Power Ratio (SLMCPR). In addition, SLO [single loop

operation] has been specified as a operating option for the

transient and accident evaluations performed as part of the cycle-

specific core reload analyses for Units 2 and 3 which ensure that

operating limit Minimum Critical Power Ratios (OLMCPRs) for the

current fuel types are established that maintain required margin to

the fuel cladding safety limit. A cycle-specific analysis with SLO

will be performed for Unit 1 prior to restart and experience

indicates similar results are expected as those for Units 2 and 3.

A review of the values used in the statistical analysis used in

the basis of the fuel cladding safety limit determined that, due to

increased uncertainties in total core flow readings and Traversing

In-Core Probe (TIP) readings during SLO, an increase in the SLMCPR

of .02 is bounding when in SLO. Therefore, while operating in

single-loop mode, an additional .02 is added to the OLMCPR which

maintains the same margin to the fuel cladding safety limit as that

established for two-loop operation. This is a conservative approach

because the two-loop transients have been shown to be more severe

than the equivalent single-loop events and, therefore, the OLMCPRs

established for two-loop operation would always be bounding. Thus,

the margin of safety for fuel clad integrity is assured and the

probability or consequences associated with reactor transients is

not increased for SLO.

SLO results in backflow through the jet pumps in the inactive

recirculation loop which perturbs the relationship between the core

flow and recirculation drive flow on which the flow biased Average

Power Range Monitor (APRM) and Rod Block Monitor (RBM) setpoint

equations are based. To compensate, the proposed TS [Technical

Specification] changes modify the setpoint equations to correct for

one-loop operation. With this adjustment, the setpoint equations

preserve the original relationship between the setpoints and the

effective recirculation drive flow such that the consequences of a

RWE [rod withdrawal event] in SLO are bounded by the cycle-specific

RWE analyses. Therefore, these changes do not increase the

probability or consequences of the RWE transient previously

evaluated.

Average Planar Linear Heat Generation Rate (APLHGR) limits are

established to ensure the acceptance criteria for fuel and Emergency

Core Cooling Systems established in 10 CFR 50.46 are met. A SLO Loss

of Coolant Accident (LOCA) analysis was performed using the SAFER/

GESTR computer code as documented in NEDC-32484P, Revision 1,

``Browns Ferry Nuclear Plant, Units 1, 2, and 3, SAFER/GESTR-LOCA,

Loss-of-Coolant Accident Analysis.''

The LOCA [loss of cooling accident] results for SLO using SAFER/

GESTR showed that, with the application of an APLHGR multiplier as

proposed in the TS change, the LOCA peak clad temperature for SLO

will always be lower than that for limiting design basis pipe break

for two-loop operation. An APLHGR multiplier of 0.9 is applicable

for all current fuel types being used. This multiplier is documented

in each cycle-specific reload analysis and included in the COLR

[core operating limits report]. NEDC-32484P Revision 1 also

concludes that the design basis accident (large breaks) are more

affected than small break sequences and, therefore, the large break

results are bounding for SLO.

The Recirculation Pump Seizure event in SLO was evaluated in

NEDO-24236 and shown to be a non-limiting event. This conclusion is

also supported by GE analyses on other BWRs.

In summary, based on the above discussion, the proposed changes

for SLO do not increase the probability or consequences of an

accident previously evaluated.

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

new or different kind of accident from any accident previously

evaluated.

Although the proposed change allows extended operation in a

configuration that was previously allowed for a limited period,

analysis has shown (as described in item A above), that operation

with one recirculation pump out-of-service is within existing

analyses based on the proposed TS requirements. Therefore, the

proposed change does not create the possibility of a new or

different kind of accident from any accident previously evaluated.

3. The proposed amendment does not involve a significant

reduction in a margin of safety.

The proposed change to operate in single-loop recirculation mode

has been analyzed in accordance with established transient and

accident methodologies, and margins of safety for the design basis

accidents and transients analyzed in Chapter 14 of the BFN UFSAR

[updated final safety analysis report] have not been significantly

reduced. The basis for this conclusion is outlined in item A above.

Therefore, the proposed change does not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

Local Public Document Room location: Athens Public Library, 405 E.

South Street, Athens, Alabama 35611

Attorney for licensee: General Counsel, Tennessee Valley Authority,

400 West Summit Hill Drive, ET l0H, Knoxville, Tennessee 37902

NRC Project Director: Frederick J. Hebdon

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

Vermont Yankee Nuclear Power Station, Vernon, Vermont

Date of amendment request: June 9, 1997

Description of amendment request: The amendment proposes to update

the Technical Specifications, Section 6.0, to add a reference to NRC-

approved methodologies which will be used to validate or generate the

operating limits in the Vermont Yankee Core Operating Limits Report.

Basis for proposed no significant hazards consideration

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

provided its analysis of the issue of no significant hazards

consideration which is presented below:

1. The proposed change will not involve any significant increase

in the probability or consequences of an accident. The change

updates the Technical Specifications to include [an] NRC approved

method reference to allow calculation of thermal hydraulic stability

limits. It does not affect plant operation and will not weaken or

degrade the facility.

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

or different kind of accident since the change is administrative. No

physical alterations of the plant, setpoint changes, or operating

conditions are proposed.

3. The proposed change will not involve a significant reduction

in a margin of safety. The change involves an update to the

Administrative Controls in Section 6.0 of the Technical

Specifications by adding a reference to NRC approved methods. This

administrative change does not alter plant safety margins.

[[Page 43378]]

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

onthis review, it appears that the three standards of 10 CFR

50.92(c) are satisfied. Therefore, the NRC staff proposes to

determine that the amendment request involves no significant hazards

consideration.

Local Public Document Room location: Brooks Memorial Library, 224

Main Street, Brattleboro, VT 05301

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

International Place, Boston, MA 02110-2624

NRC Project Director: Ronald B. Eaton, Acting

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.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,

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 amendment request: June 9, 1997

Description of amendment request: The proposed amendments authorize

a revision to the realistic dose values for the process gas system

rupture in Section 15.0 of the Byron/Braidwood (B/B) Updated Final

Safety Analysis Report (UFSAR). During preparation of a UFSAR change

package, ComEd discovered that the Final Safety Analysis Report (FSAR)

had not been updated to correct an error from the previous revision of

the dose calculation. Since the correct dose value is greater than that

previously reported, the consequences of the accident had increased,

and an unreviewed safety question resulted.

Date of publication of individual notice in Federal Register: July

10, 1997 (62 FR 37079).

Expiration date of individual notice: August 11, 1997 (as corrected

(62 FR 39282)).

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

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388

Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of amendment request: June 27, 1997, as supplemented by letter

dated July 2, 1997 The supplemental letter provided clarifying

information and did not change the initial proposed no significant

hazards consideration determination.

Brief description of amendment request: These amendments clarify,

in the technical specifications (TSs) for each unit, the methodology

used to satisfy surveillance requirements for the laboratory analysis

of activated carbon (charcoal) samples from the standby gas treatment

system (SGTS) and the control room emergency outside air supply system

(CREOASS). The specific changes are made to Sections 4.6.5.3.b.2 and

4.6.5.3.c for the SGTS and to Sections 4.7.b.2 and 4.7.2.c for the

CREOASS, to include a reference to American Society for Testing

Materials (ASTM), ``Radioiodine Testing of Nuclear-Grade Gas Phase

Adsorbents,'' ASTM D3803-79. Date of publication of individual notice

in Federal Register: July 8, 1997 (62 FR 36580)

Expiration date of individual notice: August 7, 1997

Local Public Document Room location: Osterhout Free Library,

Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701

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.

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

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

Date of application for amendments: April 14, 1997

Brief description of amendments: The amendments revise Technical

Specification (TS) 3/4.3.8, ``Feedwater/Main Turbine Trip System

Actuation Instrumentation'' by changing the minimum channels required

from three to four. This change reflects a modification that is being

installed to add an auxiliary contact to the trip system logic. In

addition, the amendments revise the TS action statement for inoperable

channels to be consistent with the Improved Standard Technical

Specifications and to account for the additional channel.

Date of issuance: July 29, 1997

Effective date: Immediately, to be implemented within 60 days.

Amendment Nos.: 119 and 104

Facility Operating License Nos. NPF-11 and NPF-18: The amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: June 18, 1997 (62 FR

33120). The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated July 29, 1997. No significant

hazards consideration comments received: No.

[[Page 43379]]

Local Public Document Room location: Jacobs Memorial Library,

Illinois Valley Community College, Oglesby, Illinois 61348

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

Buren County, Michigan

Date of application for amendment: March 27, 1997, as supplemented

July 7, 1997

Brief description of amendment: The amendment revises the Palisades

Plant license and technical specifications to reflect the licensee's

name change from ``Consumers Power Company'' to ``Consumers Energy

Company.''

Date of issuance: July 21, 1997

Effective date: July 21, 1997

Amendment No.: 176

Facility Operating License No. DPR-20: Amendment revised the

license and the technical specifications.

Date of initial notice in Federal Register: April 23, 1997 (62 FR

19828) The July 7, 1997, letter provided supplementary information

within the scope of the original application and did not change the NRC

staff's initial proposed no significant hazards considerations

determination. The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated July 21, 1997. No significant

hazards consideration comments received: No.

Local Public Document Room location: Van Wylen Library, Hope

College, Holland, Michigan 49423

Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba

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

Date of application for amendments: May 27, 1997

Brief description of amendments: The amendments delete Section

4.7.13.3.a.2 of each unit's Technical Specifications, regarding the

minimum volume and boron concentration of borated water available to

the Standby Makeup Pump of the Standby Shutdown System.

Date of issuance: July 21, 1997

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

30 days

Amendment Nos.: 160 and 152

Facility Operating License Nos. NPF-35 and NPF-52: Amendments

revised the Technical Specifications.

Date of initial notice in Federal Register: June 18, 1997 (62 FR

33121) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated July 21, 1997. No significant

hazards consideration comments received: No.

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

Black Street, Rock Hill, South Carolina 29730

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

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

Date of application for amendment: February 17, 1997, as revised

May 1, 1997.

Brief description of amendment: Changes to Technical Specification

(TS) to implement 10 CFR 50, Appendix J Option B relating to

containment leakage tests.

Date of issuance: July 24, 1997

Effective date: July 24, 1997

Amendment No.: 156

Facility Operating License No. DPR-72: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: February 28, 1997 (62

FR 9214), as superseded June 4, 1997 (62 FR 30632) The Commission's

related evaluation of the amendment is contained in a Safety Evaluation

dated July 24, 1997. No significant hazards consideration comments

received: No.

Local Public Document Room location: Coastal Region Library, 8619

W. Crystal Street, Crystal River, Florida 32629

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

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

Connecticut

Date of application for amendment: April 28, 1997

Brief description of amendment: Technical Specification (TS) 3.7.6

requires that flood protection be provided for the service water pump

cubicles and components when the water level exceeds a specific value.

The amendment (1) adds the closing of the service water pump cubicle

sump drain valves to the TS, (2) revises the wording of the action

statement to be consistent with the limiting condition for operation,

and (3) revises the associated Bases section.

Date of issuance: July 28, 1997

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

within 60 days.

Amendment No.: 144

Facility Operating License No. NPF-49: Amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: June 4, 1997 (62 FR

30636) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated July 28, 1997. No significant

hazards consideration comments received: No.

Local Public Document Room location: Learning Resources Center,

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

Norwich, Connecticut 06360, and the Waterford Library, ATTN: Vince

Juliano, 49 Rope Ferry Road, Waterford, Connecticut 06385

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

Nuclear Generating Plant, Wright County, Minnesota

Date of application for amendment: January 23, 1997, as

supplemented January 28, March 4, June 19, July 2, July 16 (2 letters),

July 21, and July 25, 1997

Brief description of amendment: The amendment documents the staff's

review and approval of the apparent unreviewed safety questions (USQs)

associated with (1) the updated analysis of the design-basis accident

(DBA) containment temperature and pressure response, and (2) the

reliance on containment pressure to compensate for the potential

deficiency in net positive suction head (NPSH) for the emergency core

cooling system (ECCS) pumps during a DBA with the worst case scenario

assumptions. The amendment also authorizes the licensee to change the

Technical Specification bases and the Updated Safety Analysis Report,

to reflect the reliance of containment pressure to compensate for the

potential deficiency in NPSH for the ECCS pumps following a DBA.

Date of issuance: July 25, 1997

Effective date: July 25, 1997. Implementation shall be as specified

in Appendix C to the license.

Amendment No.: 98

Facility Operating License No. DPR-22: Amendment revised the

license and the licensee's updated safety analysis report.

Date of initial notice in Federal Register: February 12, 1997 (62

FR 6576) The June 19, 1997, submittal, expanded the scope of the

initial submittal dated January 23, 1997, and therefore, another notice

was issued in Federal Register on June 24, 1997 (62 FR 34086). The July

2, July 16 (2 letters), July 21, and July 25, 1997, submittals provided

additional clarifying information within the scope of the application

and did not change the NRC staff's proposed no significant hazards

considerations determination that was based on the June 19, 1997,

submittal. Therefore, renoticing was not warranted. The Commission's

related evaluation of the amendment is contained in a Safety Evaluation

dated July 25, 1997. No significant hazards consideration comments

received: No.

Local Public Document Room location: Minneapolis Public Library,

[[Page 43380]]

Technology and Science Department, 300 Nicollet Mall, Minneapolis,

Minnesota 55401

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388

Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,

Pennsylvania

Date of application for amendments: June 27, 1997, as supplemented

by letter dated July 2, 1997 The supplemental letter provided

clarifying information and did not change the initial proposed no

significant hazards consideration determination.

Brief description of amendments: These amendments clarify, in the

technical specifications (TSs) for each unit, the methodology used to

satisfy surveillance requirements for the laboratory analysis of

activated carbon (charcoal) samples from the standby gas treatment

system (SGTS) and the control room emergency outside air supply system

(CREOASS). The specific changes are made to Sections 4.6.5.3.b.2 and

4.6.5.3.c for the SGTS and to Sections 4.7.b.2 and 4.7.2.c for the

CREOASS, to include a reference to American Society for Testing

Materials (ASTM), ``Radioiodine Testing of Nuclear-Grade Gas Phase

Adsorbents,'' ASTM D3803-79.

Date of issuance: July 30, 1997

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

implemented within 30 days.

Amendment Nos.: 167 and 141

Facility Operating License Nos. NPF-14 and NPF-22: The amendments

revised the Technical Specifications. Public comments requested as to

proposed no significant hazards consideration: Yes (62 FR 36580). That

notice provided an opportunity to submit comments on the Commission's

proposed no significant hazards consideration determination by July 22,

1997. No comments have been received. The notice also provided an

opportunity to request a hearing by August 7, 1997, but indicated that

if the Commission makes a final no significant hazards consideration

determination, any such hearing would take place after issuance of the

amendment. On July 9, 1997, the NRC staff issued a Notice of

Enforcement Discretion in order to delay enforcement of the current,

subject, TS requirements until the NRC could take formal action on the

July 2, 1997, application. The Commission's related evaluation of the

amendments, finding of exigent circumstances, consultation with the

State of Pennsylvania, and final no significant hazards consideration

determination are contained in a Safety Evaluation dated July 30, 1997.

Local Public Document Room location: Osterhout Free Library,

Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701

Public Service Electric & Gas Company, Docket No. 50-354, Hope

Creek Generating Station, Salem County, New Jersey

Date of application for amendment: February 11, 1997.

Brief description of amendment: This amendment changes the Hope

Creek Technical Specification (TS) Sections 3/4.8.1, ``A.C. Sources,''

6.8, ``Procedures and Programs,'' and the Bases for Section 3/4.8,

``Electrical Power Systems,'' to include: 1) the relocation of existing

surveillance requirements related to diesel fuel oil chemistry; 2) the

introduction of a new program under TS 6.8.4.e, ``Diesel Fuel Oil

Testing Program; 3) revisions to the TS Bases for Section 3/

4.8 to incorporate information associated with the TS changes; and 4)

editorial changes to implement required corrections.

Date of issuance: July 24, 1997

Effective date: As of date of issuance, to be implemented within 60

days.

Amendment No.: 100

Facility Operating License No. NPF-57: This amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: March 26, 1997 (62 FR

14469) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated July 24, 1997. No significant

hazards consideration comments received: No.

Local Public Document Room location: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey 08070

Public Service Electric & Gas Company, Docket No. 50-354, Hope

Creek Generating Station, Salem County, New Jersey

Date of application for amendment: March 3, 1997, as supplemented

by letter dated May 5, 1997

Brief description of amendment: This amendment changes Hope Creek

TSs as follows: (1) TS 3/4.3.1, ``Reactor Protection System

Instrumentation,'' TS 3/4.3.2, ``Isolation Actuation Instrumentation,''

and TS 3/4.3.3, ``Emergency Core Cooling System Actuation

Instrumentation,'' to include additional information concerning

response time testing; (2) TS 4.0.5 to reference inservice inspection

and test requirements; (3) TS 3/4.6.1, ``Primary Containment,'' and

associated Bases to reflect a design modification; (4) TS 3/4.7.7,

``Main Turbine Bypass System,'' to specify a new operability

requirement; and (5) the Bases for TS 3/4.8, ``Electrical Power

Systems.''

Date of issuance: July 24, 1997

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

60 days.

Amendment No.: 101

Facility Operating License No. NPF-57: This amendment revised the

Technical Specifications.

Date of initial notice in Federal Register: June 18, 1997 (62 FR

33131) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated July 24, 1997. No significant

hazards consideration comments received: No.

Local Public Document Room location: Pennsville Public Library, 190

S. Broadway, Pennsville, New Jersey 08070

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: February 11, 1997, as

supplemented on May 1, June 12, and July 23, 1997

Brief description of amendments: The amendments add a new Technical

Specification, 3/4.7.10, ``Chilled Water System - Auxiliary Building

Subsystem,'' and an associated Bases section to address the support

function this system provides to other necessary safety systems.

Date of issuance: July 29, 1997

Effective date: Unit 1 to be implemented prior to entering Mode 6

from the current unit outage; Unit 2 as of its date of issuance, to be

implemented within 10 days of issuance.

Amendment Nos.: 199 and 182

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

revised the Technical Specifications.

Date of initial notice in Federal Register: March 12, 1997 (62 FR

11497) The licensee's supplemental letters provided additional

information that did not affect the staff's proposed no significant

hazards consideration determination. The Commission's related

evaluation of the amendments is contained in a Safety Evaluation dated

July 29, 1997. No significant hazards consideration comments received:

No.

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

West Broadway, Salem, NJ 08079

Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear

Plant, Unit 1, Rhea County, Tennessee

Date of application for amendment: October 23, 1996, as

supplemented

[[Page 43381]]

December 11, 1996, January 31, February 10 and 24, March 11, April 4

and 11, May 28, June 26, and July 15, 1997.

Brief description of amendment: The amendment changes the Watts Bar

Nuclear Plant, Unit 1, Technical Specifications (TS) to increase the

spent fuel storage capacity from 484 fuel assemblies to 1610 fuel

assemblies and to increase the initial enrichment of the fuel to be

stored in the spent fuel storage racks from 3.5 weight percent (wt%) to

5.0 wt%. This modification also changes the center-to-center spacing of

stored fuel assemblies and reflects the use of burnup credit rack

modules to be installed peripherally along the pool walls.

The amendment, as proposed by the licensee, would also involve the

installation of spent fuel racks in the spent fuel cask pit for 225

storage spaces thus increasing the total WBN spent fuel storage

capacity to 1835 spent fuel assemblies. The licensee proposed to

provide an impact shield that would be placed over the fuel in the cask

pit when heavy loads are moved near or across the cask pit area. The

staff is continuing its review of this aspect of the licensee's

proposal. Accordingly, this amendment authorizes the reracking and

usage of the main spent fuel pool, as proposed for a total of 1610

spent fuel spaces. However, it does not authorize the installation of

storage racks or storage of spent fuel in the spent fuel cask pit. The

staff's review of that aspect of the licensee's application will be

addressed by further correspondence.

Date of issuance: July 28, 1997

Effective date: July 28, 1997

Amendment No.: 6

Facility Operating License No. NPF-90: Amendment revises the TS.

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

15733) The April 4, and 11, May 28, June 26 and July 15, 1997 letters

provided clarifying informaion that did not change the initial proposed

no significant hazards consideration determination.

The Commission's related evaluation of the amendment is contained

in an environmental assessment dated April 7, 1997, and a Safety

Evaluation dated July 28, 1997. No significant hazards consideration

comments received: None

Local Public Document Room location: Chattanooga-Hamilton County

Library, 1001 Broad Street, Chattanooga, TN 37402

Virginia Electric and Power Company, et al., Docket Nos. 50-338 and

50-339, North Anna Power Station, Units No. 1 and No. 2, Louisa

County, Virginia

Date of application for amendments: November 9, 1987, as

supplemented March 31, 1988, June 8, 1992, and February 4, 1997

Brief description of amendments: These amendments reformat the

operability and surveillance requirements for the intermediate range

channels.

Date of issuance: July 30, 1997

Effective date: July 30, 1997

Amendment Nos.: 206 and 187

Facility Operating License Nos. NPF-4 and NPF-7: Amendments revised

the Technical Specifications.

Date of initial notice in Federal Register: June 18, 1997 (62 FR

33136) The Commission's related evaluation of the amendments is

contained in a Safety Evaluation dated July 30, 1997. No significant

hazards consideration comments received: No.

Local Public Document Room location: The Alderman Library, Special

Collections Department, University of Virginia, Charlottesville,

Virginia 22903-2498

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf

Creek Generating Station, Coffey County, Kansas

Date of amendment request: February 17, 1997

Brief description of amendment: The amendment revises the technical

specifications to move Table 3.6-1, ``Containment Isolation Valves'' to

Wolf Creek Generating Station procedures. In addition, the technical

specifications have been modified to remove all references to Table

3.6-1. This change is in accordance with the guidance provided in

Generic Letter 91-08, ``Removal of Component Lists from Technical

Specifications,'' dated May 6, 1991.

Date of issuance: July 23, 1997

Effective date: July 23, 1997, to be implemented within 30 days

from the date of issuance.

Amendment No.: 108

Facility Operating License No. NPF-42: The amendment revised the

Technical Specifications and the Operating License.

Date of initial notice in Federal Register: April 23, 1997 (62 FR

19838) The Commission's related evaluation of the amendment is

contained in a Safety Evaluation dated July 23, 1997. No significant

hazards consideration comments received: No.

Local Public Document Room locations: Emporia State University,

William Allen White Library, 1200 Commercial Street, Emporia, Kansas

66801 and Washburn University School of Law Library, Topeka, Kansas

66621

Dated at Rockville, Maryland, this 6th day of August, 1997.

For the Nuclear Regulatory Commission

Jack W. Roe,

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

Reactor Regulation.

[Doc. 97-21244 Filed 8-12-97; 8:45 am]

BILLING CODE 7590-01-F