[Federal Register Volume 59, Number 166 (Monday, August 29, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-21223]


[[Page Unknown]]

[Federal Register: August 29, 1994]


-----------------------------------------------------------------------

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