[Federal Register Volume 59, Number 151 (Monday, August 8, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-19266]


[[Page Unknown]]

[Federal Register: August 8, 1994]


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NUCLEAR REGULATORY COMMISSION

[Docket Nos. 50-352 and 50-353]

 

 Philadelphia Electric Co.; Notice of Consideration of Issuance 
of Amendment 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. 
NPF-39 and NPF-85 issued to the Philadelphia Electric Company (PECO or 
the licensee) for operation of the Limerick Generating Station, Units 1 
and 2, located in Montgomery County, Pennsylvania.
    The proposed amendment request of January 14, 1994, would increase 
the storage capacity in each spent fuel pool (SFP) from their current 
2040 fuel assemblies to 4117 fuel assemblies. In addition, the proposed 
amendment would extend the ``full core reserve'' capability from year 
1998 to 2013.
    On May 6, 1994, PECO submitted another application requesting an 
interim increase in the capacity of the Unit 1 SFP, from 2040 to 2500 
fuel assemblies. PECO's submittal of June 3, 1994 supplemented their 
submittal of May 6, 1994. The Commission granted approval of the May 6, 
1994 application by Amendment No. 72 to Facility Operating License No. 
NPF-39 for the Limerick Generating Station, Unit 1, on June 30, 1994.
    Before issuance of the proposed license amendment, 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 Technical Specifications (TS) changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    Increasing the spent fuel storage capacity in each Spent Fuel 
Pool (SFP) to 4117 fuel assemblies does not increase the probability 
of occurrence of an accident. Since all fuel handling activities 
will be performed using approved procedures and compatible 
equipment, the probability of a fuel handling accident occurring is 
unchanged.
    The intermediate configuration involving the installation of the 
new maximum density racks in the Unit 2 SEP and placement of 
additional existing racks in the Unit 1 SFP will not prevent the 
ability of the Fuel Pool Cooling and Cleanup (FPCC) systems from 
adequately cooling their respective SFP. The backup cooling and 
makeup systems (i.e., Residual Heat Removal (RHR), Emergency Service 
Water (ESW), and Residual Heat Removal Service Water (RHRSW) 
systems) will continue to function as designed to provide an 
alternate source of cooling and makeup water to ensure SFP cooling 
is maintained. Increasing the spent fuel storage capacity in each 
SFP will result in a slight increase in the maximum normal decay 
heat load from 16.32  x  10\6\ Btu/hr to 18.05  x  10\6\ Btu/hr. 
This increase is due to 1) the heat load associated with a maximum 
storage capacity of 4117 fuel assemblies, 2) a 5% power rerate 
consideration (i.e., the effects of increasing the rated core 
thermal power from 3293 MWt to 3458 MWt), 3) a reduction in the 
minimum decay time until fuel movements begin, and 4) the effects of 
increasing our refueling cycles from 18-months to 24 months. Section 
9.1.3, ``Spent Fuel Pool Cooling and Cleanup,'' of NUREG-0800, 
``Standard Review Plan for the Review of Safety Analysis Reports for 
Nuclear Power Plants,'' recommends that the SFP temperature be 
maintained at or below 140 deg.F. However, due to the increase in 
the maximum normal decay heat load, and with two (2) trains of fuel 
pool cooling operating, the temperature that the SFP can be 
maintained will increase from 140 deg.F to 143 deg.F. The time 
period that two (2) trains of fuel pool cooling can not maintain the 
pool temperature below 140 deg.F is 2.5 days and the SFP temperature 
will exceed 140 deg.F approximately 160 hours after plant shutdown. 
The slight increase in SFP temperature (i.e., 140 deg.F to 
143 deg.F) is considered acceptable since the increase is small 
(i.e., 3 deg.F), and the duration in which the temperature exceeds 
140 deg.F is short (i.e., 2.5 days). In addition, during this period 
the RHR system will be available for operation to maintain the 
desired SFP temperature. The maximum decay heat load, assuming full 
core discharge and remaining cells filled, will increase from 36.4 x 
10\6\ Btu/hr to 37.6 x 10\6\ Btu hr; however, the RHR system is 
still be [sic] capable of maintaining SFP temperature less than 
140 deg.F as described in LGS Updated Final Safety Analysis Report 
(UFSAR) and supporting Safety Analysis Report provided in Attachment 
2 [See application dated January 14, 1994 for Attachment 2]. This 
increase in temperature will not increase the probability of a loss 
of fuel pool cooling accident or adversely affect the Refuel Floor 
ventilation system.
    The proposed piping modifications to the RHR system piping 
inside the Unit 2 SFP will not interfere with the RHR system's 
ability to adequately cool the SFP or to prevent siphoning of the 
SFP water.
    Movement of the Unit 2 SFP gates to the new storage location and 
installation of the new fuel storage racks will be accomplished in 
accordance with the guidance specified in NUREG-0612, ``Control of 
Heavy Loads at Nuclear Power Plants.'' Approved procedures, safe 
load paths, and single failure proof rigging will be used. 
Therefore, the probability of a heavy load drop is unchanged.
    The consequences of a Fuel Handling Accident as described in the 
LGS UFSAR is not increased since the number of fuel assemblies 
stored in a SFP is not an input to the initial conditions of the 
accident evaluation. This accident evaluates the dropping of a spent 
fuel assembly and the fuel grapple assembly into the reactor core 
during refueling operations. A drop height of 32 feet for the spent 
fuel assembly and 47 feet for the fuel grapple assembly are assumed 
and will produce the largest number of failed fuel rods. The tops of 
the new spent fuel racks are at the same level as the existing spent 
fuel racks. Since the maximum possible height a fuel assembly can be 
dropped over the SFP does not exceed 32 feet, the consequences of a 
Fuel Handling Accident will not be increased by increasing the 
number of fuel storage cells. The increase in dose estimates 
presented in the Safety Analysis Report are within 10 CFR 100 limits 
and are the result of increased fuel enrichment for power rerate and 
24-month refueling cycles, and not as a result of an increase in the 
number of fuel storage cells. These other changes are the subject of 
separate TS Change Requests that have already been submitted to the 
NRC for approval.
    The consequence of a loss of fuel pool cooling as described in 
Section 9.1.3.6 of the LGS UFSAR will not be increased. The event 
described in the UFSAR assumes that the iodine in the fuel from past 
refuelings is negligible, due to long decay time. Iodine is the 
major contributor to thyroid dose. Since the iodine in the fuel from 
past refuelings is negligible, due to the long decay time, 
increasing the number of fuel storage cells will not increase the 
dose due to the release of iodine in the SFP water resulting from 
boiling and therefore, the consequences are not increased. The time 
to boil of 13.5 hours currently specified in UFSAR bounds the time 
to boil of 9.15 hrs presented in the supporting Safety Analysis 
Report since the 13.5 hrs is for 21 days after reactor shutdown and 
the 9.15 hrs is for 7.25 days after reactor shutdown, and the decay 
heat from the newly discharged fuel decreases exponentially with 
time after plant shutdown.
    The new maximum density storage racks have been designed and 
analyzed to maintain Keff less than or equal to 0.95. The 
supporting Safety Analysis Report includes the effects of various 
anomalies such as a fuel assembly drop event, manufacturing 
tolerance variations, and abnormal location of a fuel assembly. 
Since a Keff of less than or equal to 0.95 with a confidence 
factor of 95% is maintained, the consequences of an event that would 
affect criticality control will not increase. The planned interim 
configuration of the Unit 1 pool is bounded by the current analyses 
in the UFSAR, since the rack design is unchanged.
    The new maximum density storage racks have been designed and 
analyzed to seismic Category 1 criteria and are capable of remaining 
functional during the event of a fuel assembly and fuel grapple 
assembly impacting the rack from a height of 36 inches, as described 
in the attached Safety Analysis Report [See application dated 
January 14, 1994 for Attachment 2]. Since the new maximum density 
storage racks are capable of withstanding an impact from a height of 
36 inches, the consequences of the events described in the LGS UFSAR 
which use a drop height of 16 inches, are not increased.
    Increasing the on-site storage capacity by installing additional 
storage cells will not increase the probability of a malfunction of 
the stored spent fuel based on the thermal-hydraulic analysis 
presented in the supporting Safety Analysis Report [See application 
dated January 14, 1994 for Attachment 2] which concludes that 
sufficient cooling exists with 4117 fuel assemblies in a SFP. As for 
fuel criticality, the determination is based on the criticality 
analysis documented in the supporting Safety Analysis Report which 
confirms that the stored fuel assemblies will remain sub-critical 
under normal and abnormal conditions.
    Increasing the on-site storage capacity by installing additional 
storage cells will not increase the probability of a malfunction of 
the SFP liner based upon the SFP structural analysis as documented 
in the supporting Safety Analysis Report which indicates that 
adequate margin exists to prevent overstressing of the SFP liner.
    Increasing the on-site capacity by installing addition[al] 
storage cells will not increase the probability of a malfunction of 
the SFP structure. This is based upon the SFP structural analysis as 
documented in the supporting Safety Analysis Report which confirms 
that the SFP structure still has adequate margin to prevent 
overstressing and meets the code requirements for the LGS.
    Increasing the on-site storage capacity by installing additional 
storage cells will not increase the probability of a malfunction of 
the spent fuel storage racks based on the seismic/structural 
analysis documented in the supporting Safety Analysis Report which 
concludes that interaction of racks during a seismic event will not 
result in loss of spent fuel storage racks' ability to function. The 
planned relocating the storage location of the SFP gates will not 
increase the probability of a malfunction of the SFP gates since, 
while being store, the SFP gates do not perform a safety function. 
The hangers used to secure the SFP gates will be designed/installed 
to the same requirements as the existing hangers.
    Increasing the on-site spent fuel storage capacity will not 
increase the probability of a malfunction of the Fuel Pool Cooling 
and Cleanup (FPCC) system. The only impact on the FPCC system of 
increasing the spent fuel storage capacity will be a slight increase 
in fluid temperature (i.e., 140 deg.F to 143 deg.F) which is within 
the design temperature of the system (i.e., 150 deg.F) as described 
in the LGS UFSAR.
    Modifying the RHR piping in the Unit 2 SFP such that it will not 
interfere with increased fuel storage will not increase the 
probability of a malfunction of the RHR system since the RHR 
system's ability to cool the SFP and to prevent siphoning of the SFP 
water will remain unchanged. Only the RHR discharge piping inside 
the SFP will be modified. the proper flow pattern will be maintained 
and net postivie suction head requirements will be unaffected.
    The probability of a malfunction of fuel handling equipment will 
not be increased since increasing the on-site storage capacity does 
not affect fuel handling equipment.
    Increasing the on-site spent fuel storage capacity does not 
increase the consequences of a spent fuel assembly failure since the 
failure of one assembly will not result in additional spent fuel 
assembly failures.
    Increasing the on-site spent fuel storage capacity does not 
increase the consequences of a loss of fuel pool cooling as 
described in Section 9.1.3.6 of the LGS UFSAR which evaluated the 
radiological affects due to thyroid dose. Iodine is the major 
contributor to thyroid dose. The iodine in the fuel from past 
refuelings is negligible, due to the long decay time. Since the 
release of iodine resulting from the SFP water boiling is entirely 
due to the freshly discharged fuel, the consequences of reracking 
the SFPs are unchanged from that previously evaluated. The 
evaporation rate will increase due to higher decay heat load. 
However, since the time to boil is 9.15 hours, as discussed 
previously, adequate time exists to align the alternate makeup water 
sources (e.g., RHR, Emergency Service Water (ESW), and Residual Heat 
Removal Service Water (RHRSW) systems) to maintain SFP water level 
and therefore, the consequences are not increased.
    Increasing the on-site storage capacity will not increase the 
consequences of spent fuel storage rack failure, since both the new 
maximum density racks and the existing racks have been designed/
qualified to limit the consequences of a failure. A failure of or 
damage to one (1) storage rack will not result in failure or damage 
to another storage rack.
    Increasing the on-site storage capacity will not increase the 
consequences of a failure of the SFP gates or SFP liner since the 
design of the SFP to maintain adequate water level and the available 
makeup capacity are unaffected.
    Increasing the on-site storage capacity will not increase the 
consequences of the failure of fuel handling equipment since the 
maximum expected number of fuel rods damaged by a fuel handling 
equipment failure remains as evaluated in the LGS UFSAR.
    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.
    Increasing the spent fuel storage capacity in each of the SFPs 
at LGS to a maximum of 4117 fuel assemblies as analyzed in the 
attached Safety Analysis Report [See application dated January 14, 
1994 for Attachment 2] will not create the possibility of an 
accident of a different type. The SFP configurations have been 
analyzed for reactivity/criticality effects, thermal/seismic-
structural effects, radiological effects, and thermal-hydraulic 
effects. Sine the increase in storage capacity is achieved by 
installation of additional storage racks which are passive 
components, the possibility of creating a new accident does not 
exist.
    No new operating schemes or active equipment types will be 
required to store additional fuel bundles in the SFP. Therefore, the 
possibility of a different type of malfunction occurring is not 
created.
    Therefore, the proposed TS changes do not create [the] 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    Since the existing TS limits for fuel handling interlocks, heavy 
loads restrictions, water coverage over irradiated fuel, and fuel 
sub-criticality will be maintained, the margin of safety will not be 
reduced.
    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.
    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 P-223, Phillips Building, 7920 Norfolk Avenue, 
Bethesda, 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 requests for hearing and petitions for leave to 
intervene is discussed below.
    By September 7, 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 participates 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 Pottstown Public Library, 500 High 
Street, Pottstown, Pennsylvania 19464. 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 Charles L. Miller: petitioner's name and 
telephone number, date petition was mailed, plant name, and publication 
date and page number of this Federal Register notice. A copy of the 
petition should also be sent to the Office of the General Counsel, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555, and to J.W. 
Durham, 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).
    The Commission hereby provides notice that this is a proceeding on 
an application for a license amendment falling within the scope of 
Section 134 of the Nuclear Waste Policy Act of 1982 (NWPA), 42 U.S.C. 
10154. Under Section 134 of the NWPA, the Commission, at the request of 
any party to the proceeding, must use hybrid hearing procedures with 
respect to ``any matter which the Commission determines to be in 
controversy among the parties.'' The hybrid procedures in Section 134 
provide for oral argument on matters in controversy, preceded by 
discovery under the Commission's rules, and the designation, following 
argument, of only those factual issues that involve a genuine and 
substantial dispute, together with any remaining questions of law, to 
be resolved in an adjudicatory hearing. Actual adjudicatory hearings 
are to be held on only those issues found to meet the criteria of 
Section 134 and set for hearing after oral argument.
    The Commission's rules implementing section 134 of the NWPA are 
found in 10 CFR Part 2, Subpart K, ``Hybrid Hearing Procedures for 
Expansion of Spent Nuclear Fuel Storage Capacity at Civilian Nuclear 
Power Reactors'' (published at 50 FR 41662, October 15, 1985) to 10 CFR 
2.1101 et seq. Under those rules, any party to the proceeding may 
invoke the hybrid hearing procedures by filing with the presiding 
officer a written request for oral argument under 10 CFR 2.1109. To be 
timely, the request must be filed within 10 days of an order granting a 
request for hearing or petition to intervene. (As outlined above, the 
Commission's rules in 10 CFR Part 2, Subpart G, and 2.714 in 
particular, continue to govern the filing of requests for a hearing or 
petitions to intervene, as well as the admission of contentions.) The 
presiding officer shall grant a timely request for oral argument. The 
presiding officer shall grant an untimely request for oral argument 
only upon showing of good cause by the requesting part for the failure 
to file on time and after providing the other parties an opportunity to 
respond to the untimely request. If the presiding officer grants a 
request for oral argument, any hearing held on the application shall be 
conducted in accordance with hybrid hearing procedures. In essence, 
those procedures limit the time available for discovery and require 
that an oral argument be held to determine whether any contentions must 
be resolved in adjudicatory hearing. if no party to the proceedings 
requests oral argument, or if all untimely requests for oral argument 
are denied, then the usual procedures in 10 CFR Part 2, Subpart G, 
apply.
    For further details with respect to this action, see the 
application for amendment dated January 14, 1994, 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 Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.

    Dated at Rockville, Maryland, this 1st day of August 1994.

    For the Nuclear Regulatory Commission.
Frank Rinaldi,
Project Manager, Project Directorate I-2, Division of Reactor 
Projects--I/II, Office of Nuclear Reactor Regulation.
[FR Doc. 94-19266 Filed 8-5-94; 8:45 am]
BILLING CODE 7590-01-M