Philadelphia Electric Co.; Notice of Consideration of Issuance of Amendments to Facility Operating License, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing

NUCLEAR REGULATORY COMMISSION

[Docket Nos. 50-277 and 50-278]

Philadelphia Electric Co.; Notice of Consideration of Issuance of

Amendments to Facility Operating License, Proposed No Significant

Hazards Consideration Determination, and Opportunity for a Hearing

The U.S. Nuclear Regulatory Commission (the Commission) is

considering issuance of an amendment to Facility Operating License Nos.

DPR-44 and DRP-56 issued to the Philadelphia Electric Company (the

licensee) for operation of the Peach Bottom Atomic Power Station, Units

2 and 3, located in York County, Pennsylvania.

The proposed amendments would revise the facility operating license

and Appendix A and B of the operating license to change the maximum

core power limit from 3293 MWt to 3458 MWt.

Before issuance of the proposed license amendments, the Commission

will have made findings required by the Atomic Energy Act of 1954, as

amended (the Act) and the Commission's regulations.

The Commission has made a proposed determination that the amendment

request involves 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. 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 OL changes do not involve a significant

increase in the probability or consequences of an accident

previously evaluated.

The proposed power rerate imposes only minor increases in the

plant operating conditions. Plant systems, components, and

structures have been verified to be capable of performing their

intended functions under rerated conditions. When necessary, some

components will be modified or replaced prior to implementation of

the Power Rerate Program to accommodate the revised operating

conditions. No new component or system interactions that could lead

to an accident are created. As discussed below, no transient events

result in a new sequence of events which could lead to a new

accident scenario.

Anticipated Transients Without Scram (ATWS) Analysis

The changes to plant parameters are consistent with the results

in NEDC-3198P, ``Generic Evaluations of General Electric Boiling

Water Reactor Power Uprate,'' dated July 1991. Therefore, the

response to an ATWS event at rerated power will be consistent with

the generic response and is acceptable.

ECCS-LOCA Analysis

The current ECCS-LOCA performance analysis already bounds the

rerated power conditions. The peak clad temperature for rerated

conditions is 1,516 deg.F which is below the 10 CFR 50.46 required

limit of 2,200 deg.F. Therefore, the analysis demonstrates that

PBAPS, Units 2 and 3 will continue to comply with 10 CFR 50.46 and

10 CFR 50, Appendix K at rerated conditions.

Abnormal Operating Transient Analysis

The results of the evaluation of transients indicate that the

margin to the Safety Limit Minimum Critical Power Ratio (SLMCPR) is

unchanged for the 8x8 array fuel types such as the GE9 product line

currently in the Unit 2 and Unit 3 cores, and will increase by 0.01

for the GE11 fuel design. The fuel thermal-mechanical limits at

power rerate conditions are within the specific design criteria for

the GE fuels currently loaded in the PBAPS Unit 2 Cycle 10 core.

Also, the power-dependent and flow-dependent MCPR and Maximum

Average Planar Linear Heat Generation Rate (MAPLHGR) limits

developed as part of the core performance improvement program are

applicable to rerated conditions. The peak PRV bottom head pressure

is still within the ASME requirement for RPV overpressure

protection.

The analysis performed focused on the most limiting transient

events in each disturbance category selected specifically for the

power rerated evaluations. The results demonstrate that PBAPS Units

2 and 3 core thermal power output can be safely increased to the

power rerate level without significant impact on the plant safety

during a postulated transient event.

(a) Events Resulting in a Nuclear System Pressure Increase

(i) Main Generator Load Rejection with No Steam Bypass

At rerated conditions, the fuel transient thermal and mechanical

overpower results remain below the NRC accepted design criteria.

(ii) Main Turbine Trip with No Steam Bypass

The fuel transient thermal responses are less severe than for

the Generator Load Rejection event. Therefore, at rerated

conditions, this event remains bounded by the Generator Load

Rejection event.

(iii) Main Steam Isolation Valve Closure, Flux Scram

The peak RPV bottom head pressure for rerated conditions is

slightly higher than the RPV bottom head pressure at current rated

conditions due to the higher initial system pressure. However, the

resultant pressure is still below the ASME overpressure limit of

1,375 psig by a margin of 68 psi.

(b) Events Resulting in a Reactor Vessel Water Temperature Decrease

(i) Inadvertent HPCI Actuation

For the condition analyzed, both the high water level setpoint

and the high RPV steam dome pressure SCRAM setpoint are not reached.

Based on the peak average fuel surface heat flux results, the HPCI

actuation event will be bounded by the limiting pressurization event

with respect to delta Critical Power ratio ([delta] CPR)

considerations. In addition, the fuel transients thermal and

mechanical overpower limits remain within the allowable NRC accepted

design values.

(ii) Feedwater Controller Failure-Maximum Demand

The [delta] CPR calculated for this event at rerated conditions

is about 0.01 higher than the corresponding value for the current

rated power. However, the trend for the Feedwater Controller

Failure-Maximum Demand event is consistent with the analysis for the

current rated power. This event continues to be the limiting event

at the low core flow condition and is bounded by the limiting

Generator Load Rejection event. The fuel thermal margin results are

within the acceptable limits for the fuel type analyzed.

(iii) Loss of Feedwater Heating

The [delta] CPR for this event at the rerated conditions is

bounded by the result estimated for this event at the current rated

power level, and remains significantly less than the cycle operating

MCPR limit. Because of the round-off process, there is no change

between the [delta] CPR results for high and low core flow

conditions. However, the results at low core flow conditions are

actually slightly higher than for the high core flow condition

because of the increased inlet coolant subcooling into the reactor

core. The calculated thermal and mechanical overpower limits for

this event at power rerate conditions also meet the fuel design

criteria.

(c) Event Resulting in a Positive Reactivity Insertion

(i) Rod Withdrawal Error (RWE)

The [delta] CPR calculated for this event at rerated conditions

is slightly less than the value for this event at current rated

power and is bounded by the generic RWE limits of 0.13 based on

implementation of the APRM-Rod Block Monitor TS (ARTS) changes.

Therefore, the generic ARTS-based RWE analysis [delta] CPR result is

verified for applicability to PBAPS power rerate conditions.

(d) Event Resulting in a Reactor Vessel Coolant Inventory Decrease

(i) Loss of Feedwater Flow

This transient event does not pose any direct threat to the fuel

in terms of a power increase from the initial conditions. However,

it is included in the power rerate evaluation to provide assurance

that sufficient water make-up capability is available to keep the

core covered when all normal feedwater is lost.

The generic analysis results in NEDC-31984P, ``Generic

Evaluations of General Electric Boiling Water Reactor Power

Uprate,'' dated July 1991, show that at power rerate conditions, the

minimum water level is reduced by about 1.5 feed from that

previously calculated for current rated power, but a large amount of

water, more than 5 feet, remains above the top of the active fuel.

The sensed water level outside of the core shroud has also been

checked to show adequate operational flexibility exists for setting

the Level 1 RPV water level setpoint so that it is not expected to

be reached even in the conservative case of a HPCI failure.

Therefore, PBAPS, Units 2 and 3 will maintain adequate reactor water

level during a postulated Loss of Feedwater Flow event at power

rerate conditions.

(e) Event Resulting in a Core Coolant Flow Decrease

(i) Recirculation Pump Seizure

The recirculation pump seizure assumes instantaneous stoppage of

the pump motor shaft of one recirculation pump. As a result, the

core flow decreases rapidly. The RPV water level swell due to the

rapid core flow reduction reaches the high RPV water level setpoint,

causing a feedwater pump strip, a turbine trip and subsequently a

reactor SCRAM on turbine stop values closure. The peak neutron flux

and average fuel surface heat flux do not increase significantly

above the initial conditions; therefore, no impact on the fuel

thermal margin is postulated to occur.

(f) Event Resulting in a Core Coolant Flow Increase

(i) Recirculation Flow Controller Failure Increasing Flow

The results of this transient for PBAPS, Units 2 and 3 power

rerate remain non-limiting as compared with other more severe

pressurization events.

(g) Performance Improvements

(i) Main Turbine Bypass Out-of-Service

The main turbine steam bypass out-of-service condition is

included in the input assumptions used in the Abnormal Operating

Transient Occurrences analyses for power rerate application. The

transient analyses results at power rerate conditions reflect the

plant response accounting for this condition.

(ii) Single Loop Operation (SLO)

The safety analysis for rerated conditions shows that the SLO

mode is valid for power rerate conditions and remains unchanged from

the current rated power conditions.

(iii) Final Feedwater Temperature Reduction

Final Feedwater Temperature Reduction is a cycle extension mode

of operation, used in conjunction with increased core flow (ICF) at

the end of a normal operating cycle. The analyses show that for a

temperature reduction up to 55 deg.F, this mode of operation is

applicable for operation of PBAPS, Units 2 and 3 at the power rerate

conditions.

(h) Other evaluations

These evaluations included the effect of power rerate on the

radiological consequences of accidents presented in UFSAR

Subsections 5.2, 14.6 and 14.9. The following bounding analyses were

performed: (1) Loss-of-Coolant Accident (LOCA); (2) Main Steam Line

Break (MSLB) Accident; (3) Fuel Handling Accident; (4) Control Rod

Drop Accident; and (5) Instrument Line Break Accident.

The analyses shows the offsite radiological consequences for the

bounding accidents increase, but remain well within the guidelines

of 10 CFR 100 as discussed in the UFSAR Section 14.9 and the NRC

Safety Evaluation Reports for PBAPS, Units 2 and 3. In general,

offsite doses are expected to increase proportionally with reactor

power. However, a direct comparison between the original analyses

and rerate values has limited meaning because the original analyses

could not be fully reconstituted. For the fuel handling accident,

control rod drop accident, and instrument line break accident, the

offsite doses increase by less than 1 rem. For the MSLB accident,

the whole body dose remains less than 1 rem and the thyroid dose

increases by only 3% from 85 rem to 88 rem. For the LOCA, a re-

evaluation of the original analysis was performed. The resultant

thyroid dose increased by 19% from 201 rem to 239 rem; however, only

about 3% of the increase is due to rerated conditions and 16% due to

changes in the analysis model reconstitution. Whole body dose

increases slightly to 3.9 rem.

Accident radiological consequences in the Control Room and

Technical Support Center (TSC) were also evaluated. The results show

doses are well below the 30-day limit of GCC 19 of Appendix A to 10

CFR 50 (i.e., 5 rem whole body and 30 rem thyroid). A re-evaluation

of the original analysis was performed. The highest dose consequence

is from a main steam line break which results in a dose of 18 rem

thyroid compared to 1.5 rem in the UFSAR. However, only about 3% of

this increase is due to rerated conditions and 16% is due to

analysis model reconstitution. All whole body doses are less than 1

rem.

An evaluation was performed to address the impact of power

rerate on accident mitigation features, structures, systems, and

components within the balance of plant. The results are as follows:

--Auxiliary systems such as primary containment chilled water,

building Heating, Ventilation, and Air Conditioning (HVAC) systems,

reactor building closed loop cooling, service water and emergency

service water, high pressure service water, spent fuel pool cooling,

process auxiliaries such as instrument air and makeup water and the

post-accident sampling system were confirmed to operate acceptably

under normal and accident conditions at rerated conditions.

--Combustible gas control systems were confirmed to be capable of

maintaining oxygen concentrations inside the primary containment

within limits under post accident conditions after implementation of

the Power Rerate Program.

--The secondary containment and standby gas treatment system were

confirmed to be able to adequately contain, process, and control the

release of normal and post-accident levels of radioactivity at

rerated conditions.

--Instrumentation was reviewed and confirmed to be capable of

performing its control and monitoring functions under rerated

conditions.

--Electric power systems including the turbine generator and

switchgear components were verified as being capable of providing

the electrical load as a result of the rerated power levels. No

safety-related electrical loads were affected which would impact the

emergency diesel generators.

--Piping systems were evaluated for the effect of operation at

higher power levels, including transient loadings. The evaluation

confirmed that with few exceptions piping and supports are adequate

to accommodate the increased loadings resulting from operation at

rerated power conditions. In a few cases, piping supports will be

modified to accept the higher forces due to rerated conditions.

--The effect of rerated conditions on high energy line break (HELB)

for all NSSS and BOP systems were evaluated. The evaluation

confirmed structures, systems, and components important to safety

are capable of accommodating the effects of jet impingement and

blowdown forces and the environmental effects resulting from HELB

events at rerated conditions.

--Control room habitability was evaluated. Post-accident Control

Room and TSC doses at rerated conditions were confirmed to be within

the limits of GDC 19 of 10 CFR 50, Appendix A.

--Doses for normal operation at rerated conditions were reviewed and

confirmed to remain within the limits of 10 CFR 20 and 10 CFR 50,

Appendix I. The impact on post-accident sampling activities and

post-accident access to vital areas was also confirmed to be

acceptable.

--The environmental qualification of equipment important to safety

was evaluated for the impact of normal and accident operating

conditions at rerated power levels. The majority of equipment

remains qualified for the new conditions. For equipment not

qualified corrective actions will be taken to ensure the plant

equipment will perform their intended functions under rerated

conditions. No new equipment will be added for power rerate which

would increase the potential for component failure. The Preventative

[Preventive] Maintenance Program (PMP) is not power dependent and

will continue to provide for equipment repair or replacement at

rerated power conditions.

--The impact of operation at rerated power levels was evaluated for

Station Blackout and fire safe shutdown area heat-up concerns. The

evaluation confirmed there is no adverse impact from rerated

conditions on the ability of the plant to achieve safe shutdown

under these conditions.

The consequences of all transients and special events (i.e.,

ATWS and Station Blackout) remain within NRC accepted criteria for

rerated conditions. Concurrent malfunctions assumed to occur during

accidents have been accounted for in the safety analyses for rerated

conditions. The consequences of these equipment malfunctions do not

change with implementation of the Power Rerate Program.

All equipment ``Important to Safety'' is capable or will be

modified/replaced to be capable of performing its intended function.

The availability of redundant systems to provide safety functions in

the event of component malfunction is not impacted as a result of

rerated conditions.

Furthermore, the impact of power rerate on the consequences of

abnormal transients and accident conditions which are a result of

component malfunctions has been shown to be acceptable.

The probability (i.e., frequency of occurrence) of DBAs

occurring is not affected by the increased power level, as the

applicable regulatory criteria established for plant equipment

(e.g., ANSI Standard B31.1, ASME code, NRC Regulatory Guides) will

still be followed as the plant is operated at the rerated power

level. Reactor SCRAM setpoints will be established such that there

is no significant increase in scram frequency due to rerated

conditions. No new challenges to safety-related equipment will

result from power rerate.

The changes in consequences of hypothetical accidents which

would occur from 102% of the rerated power, compared to those

previously evaluated, are in all cases not significant, because the

accident evaluations from a power rerate to 105% of original rated

power will not result in exceeding the applicable NRC approved

acceptance limits. The spectrum of hypothetical accidents and

transients has been investigated, and have been determined to meet

the current regulatory criteria for PBAPS, Units 2 and 3 at rerated

conditions. The offsite doses resulting from DBAs are calculated to

increase only a few percent (i.e., approximately 3%) because of the

rerated power level and remain below 10 CFR 100 limits. In the area

of core design, the fuel operating limits will still be met at the

rerated power level, and fuel reload analyses will show plant

transients meet the criteria accepted by the NRC as specified in

NEDO-24011, ``GESTAR II.''

Challenges to fuel or ECCS performance were evaluated and shown

to still meet the criteria of 10 CFR 46 and 10 CFR 50, Appendix K.

Challenges to the containment have been evaluated and still meet 10

CFR 50, Appendix A GDC 38, Long Term Cooling, and GDC 50,

Containment. Radiological Release events have been evaluated and

shown to meet the guidelines of 10 CFR 100. Therefore, the proposed

OL changes do not involve a significant increase in the probability

or consequences of an accident previously evaluated.

(2) The proposed OL changes do not create the possibility of a

new or different kind of accident from any accident previously

evaluated.

All actions to ensure that safety-related structures, systems,

and components will remain within their design allowable values and

ensure they can perform their intended functions under rerated

conditions will be taken prior to implementation of power rerate.

Power rerate does not increase challenges to or create any new

challenge to safety-related equipment or other equipment whose

failure could cause an accident. No new equipment is added as a

result of implementing the Power Rerate Program which could create

the possibility of a new type of accident. In addition, power rerate

does not create any new sequence of events or failure modes that

lead to a new type of accident.

No new operating mode, safety-related equipment lineup, accident

scenario, or equipment failure mode was identified as resulting from

the implementation of the Power Rerate Program. The full spectrum of

accident considerations defined in NRC Regulatory Guide 1.70 have

been evaluated for rerated conditions and no new or different kind

of accident has been identified. Implementation of the Power Rerate

Program uses already-developed technology and applies it within the

capabilities of already existing plant equipment in accordance with

presently existing regulatory criteria to include applicable NRC

approved codes, standards, and methods. GE has designed BWRs of

higher power levels than the rerated power of any of the currently

operating BWR fleet and no new power dependent accidents have been

identified.

Therefore, the proposed OL changes do not create the possibility

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

evaluated.

(3) The proposed OL changes do not involve a significant

reduction in a margin of safety.

Power rerate will not involve a significant reduction in a

margin of safety, as plant equipment and reactions to transients and

hypothetical accidents will not result in exceeding the presently

approved NRC acceptance limits.

For systems addressed in the TS Sections 2.1, 2.2, 3.1, 3.2,

3.4, 3.5, 3.6, and 3.7 (i.e., RPS, Protective Instrumentation, SLCS,

HPCI, RCIC, Primary System Boundary and Containment Systems) all

components will be operable and capable of performing their intended

functions under power rerate conditions such that the existing

margin of safety is not impacted.

For TS Bases 3.7.A and 4.7.A, the impact of rerated conditions

affects LOCA offsite radiological consequences discussed in that

section. A re-evaluation of the original analysis was performed. The

resultant offsite thyroid dose increased by 19% from 201 rem to 239

rem; however, only about 3% of the increase is due to rerated

conditions and 16% is due to the analysis model reconstituted. This

preserves adequate margin between expected offsite doses and 10 CFR

100 guidelines.

The events (i.e., transients and accidents) from the TS Bases

(e.g. TS Bases 2.1, 3.1) were evaluated for rerated conditions.

Although some changes to the TS are required for power rerate, no

NRC acceptance limit will be exceeded. Therefore, the margins of

safety to the safety limits and other TS limits will be maintained.

Therefore, the proposed OL changes do not involve a significant

reduction in a margin of safety.

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

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

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

amendment request involves no significant hazards consideration.

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.

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 Rules Review and

Directives Branch, Division of Freedom of Information and Publications

Services, Office of Administration, U.S. Nuclear Regulatory Commission,

Washington, DC 20555, 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

20555.

The filing of request for hearing and petitions for leave to

intervene is discussed below.

By September 28, 1994, 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 20555 and at the local

public document room located at the Government Publications Section,

State Library of Pennsylvania, (Regional Depository) Education

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

Pennsylvania 17105. 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 hearing or an

appropriate order.

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

shall set forth with particularity the interest of the petitioner in

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

the proceeding. The petition should specifically explain the reasons

why intervention should be permitted with particular reference to the

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

Act to be made 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, 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 20555, by the

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

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

the Commission by a toll-free telephone call to Western Union at 1-

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

operator should be given Datagram Identification Number N1023 and the

following message addressed to Mohan C. Thadani, Acting Director,

Project Directorate I-2: petitioner's name and telephone number, date

petition was mailed, plant name, and publication date and page number

of this Federal Register notice. A copy of the petition should also be

sent to the Office of General Counsel, U.S. Nuclear Regulatory

Commission, Washington, DC 20555, and to J.W. Durham, Sr., Esquire, Sr.

V.P. and General Counsel, Philadelphia Electric Company, 2301 Market

Street, Philadelphia, Pennsylvania 19101, attorney for the licensee.

Nontimely filings of petitions for leave to intervene, amended

petitions, supplemental petitions and/or requests for hearing will not

be entertained absent a determination by the Commission, the presiding

officer or the presiding Atomic Safety and Licensing Board that the

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

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 dated June 23, 1993, which is available for

public inspection at the Commission's Public Document Room, the Gelman

Building, 2120 L Street, NW., Washington, DC 20555 and at the local

public document room located at the Government Publications Section,

State Library of Pennsylvania, (Regional Depository) Education

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

Pennsylvania 17105.

Dated at Rockville, Maryland, this 23rd day of August 1994.

For The Nuclear Regulatory Commission

Joseph W. Shea,

Project Manager, Project Directorate I-2, Division of Reactor

Projects--I/II, Office of Nuclear Reactor Regulation.

[FR Doc. 94-21223 Filed 8-26-94; 8:45 am]

BILLING CODE 7590-01-M