[Federal Register Volume 62, Number 63 (Wednesday, April 2, 1997)]
[Notices]
[Pages 15733-15737]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-8401]
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NUCLEAR REGULATORY COMMISSION
[Docket No. 50-390]
Watts Bar Nuclear Plant, Unit 1; 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 No.
NPF-90, issued to the Tennessee Valley Authority (TVA or the licensee),
for operation of the Watts Bar Nuclear Plant (WBN), Unit 1 located in
Rhea County, Tennessee. This Notice supersedes a Notice placed in the
Federal Register on March 26, 1997 (62 FR 14469) on this matter.
The proposed amendment would revise the Watts Bar Nuclear Plant
(WBN) Unit 1 Technical Specifications to increase the enrichment and
storage capacity of the spent fuel pool racks. The proposed
modification increases the WBN spent fuel storage capacity from 484
fuel assemblies to 1835 fuel assemblies. The initial enrichment of the
fuel to be stored in the spent fuel storage racks will be increased
from 3.5 weight percent (wt%) to 5.0 wt%. This modification would also
change the spacing of stored fuel assembly center-to-center spacing
from a nominal 10.72 inches to 10.375 inches in 24 PaR flux trap rack
modules and 8.972 inches in ten smaller burnup credit rack modules to
be installed peripherally along the south and west pool walls and in a
single 15 x 15 burnup credit rack to be installed in the cask pit.
In addition to the above proposed revisions, two limiting
conditions for operation will be added to require that the combination
of initial enrichment and burnup of each spent fuel assembly to be
stored is in the acceptable region and to require boron concentration
of the cask pit to be greater than or equal to 2000 parts per million
(ppm) during fuel movement in the flooded cask pit. As an added
protection to the fuel stored in the cask pit area, the Technical
Requirements Manual (TRM) is being revised to require that an impact
shield be in place over the fuel when heavy loads are moved near or
across the cask pit area.
The WBN Unit 1 Technical Specification Bases and the TRM would be
revised to support these changes.
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:
The Nuclear Regulatory Commission has provided standards for
determining whether a significant hazards consideration exists (10
CFR 50.92(c)). A proposed amendment to an operating license for a
facility involves no significant hazards consideration if 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. Each standard is discussed below for the proposed
amendment.
(1) Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The following potential scenarios were considered:
1. A spent fuel assembly drop.
2. Drop of the transfer canal gate or the cask pit divider gate.
3. A seismic event.
4. Loss-of-cooling flow in the spent fuel pool.
5. Installation activities.
The effect of additional spent fuel pool storage cells fully
loaded with fuel on the
[[Page 15734]]
first four potential accident scenarios listed above has been
considered. It was concluded that after installation activities have
been completed, the presence of additional fuel in the pool does not
increase the probability of occurrence of these four events. Also,
based on evaluations of bulk pool temperature, rack seismic
responses, and refueling accidents, it is reasonable to conclude
that there is no significant increase in the consequences of these
events after installation is complete (See Reference 1). During the
installation activities, the following considerations support a
conclusion that neither the probability or consequences of these
four scenarios would be significantly increased.
A spent fuel assembly cannot be dropped during installation of
the 24 Programmed and Remote System Corporation (PaR) flux trap rack
modules because this activity will take place before the end of
operating cycle one and there will be no spent fuel in the WBN pool
to be moved or shuffled. Before installing the ten smaller burnup
credit racks in the pool, some fuel will be moved to create a three
foot lateral free zone clearance from stored fuel. This would
involve a one-time movement of an estimated maximum of 225 fuel
assemblies, which is less tha[n] half the fuel movements during one
refueling outage. This does not significantly increase the
probability of dropping a fuel assembly, particularly when the many
administrative controls and physical limitations imposed on fuel
handling operations are considered. The fuel handling system
consists of equipment and structures utilized for safely
implementing refueling operations in accordance with requirements of
General Design Criteria 61 and 62 of 10 CFR 50, Appendix A. The
radiological dose consequences of dropping a 5.0 wt% fuel assembly
are different from the previous FSAR [Final Safety Analysis Report]
evaluation for the 3.5 wt% fuel assembly. The Beta and Gamma doses
decrease and the maximum thyroid dose increase is less than 9%.
Therefore, the change in calculated dose values is insignificant and
remains well within regulatory guidelines.
It may be necessary to move the transfer canal gate and the cask
pit divider gate between their gated and stored positions during
installation of the burnup credit ``baby'' rack modules along the
south and west walls. During rack installation, the previously
mentioned three foot lateral free zone clearance to stored fuel
would exist. Therefore, no heavy load would be carried directly over
irradiated fuel during installation of the racks. There are numerous
design features which comply with NUREG-0612 to preclude these gates
from dropping on spent fuel. These features include design of the
lifting devices, design of the crane, and use of written procedures.
Also, the evaluation results for a gate drop on the racks indicates
that permanent damage to a fuel storage cell is limited to a maximum
depth of less than six inches below the top of the rack with no
effect on the subcriticality of fuel stored in adjacent cells. Based
on the foregoing, it is reasonable to conclude that gate handling
during the installation of the ``baby'' racks would not involve a
significant increase in the probability or consequences of an
accident.
The probability of a seismic event is not related to
installation activities. The worst consequence resulting from a
seismic event during installation activities would occur during
handling of a rack. The consequences would be insignificant because
the Auxiliary Building crane is seismically qualified and both
handling equipment and operations meet the criteria of NUREG-0612.
Nevertheless, if the seismic event resulted in a rack drop, the
consequences are insignificant, i.e., localized damage to the pool
liner and a minor leak rate which would be small in comparison to
available installed makeup capacity. The cooling and shielding of
the spent fuel would remain unaffected. Also the racks being moved
are empty during installation and therefore, the criticality
consequences of seismic events are bounded by evaluations for loaded
racks.
Rack installation activities cannot cause an accidental loss-of-
cooling flow in the spent fuel pool. The vital components of the spent
fuel pool cooling and cleanup system (SFPCCS) are not located proximate
to the pool installation activities. Coolant flow may be deliberately
curtailed to facilitate installation of the ``baby'' racks directly
beneath the discharge piping in the southwest corner of the pool. The
effects of such an action would be readily minimized and made
inconsequential during the detailed installation planning phase by
selecting a time when decay heat input from stored fuel is relatively
constant. Also careful preplanning of the work would minimize out-of-
service time and provide for intermittent coolant flow restart, if
necessary, to maintain acceptable bulk coolant temperatures. Similarly,
the effect of an independently initiated loss-of-coolant flow incident
on reracking activities can be easily accommodated by stopping work, as
necessary, to mitigate any adverse effects on the installation process.
The consequences of loss-of-cooling flow in the spent fuel pool during
installation are bounded by the analysis in Chapter 5 of the report
which includes the situation in which ``baby'' racks and the 15 x 15
cask pit rack are installed, and the pool is filled to capacity with
spent fuel.
With regard to the actual installation activities, the existing
WBN TRM prohibits loads in excess of 2059 pounds from travel over
fuel assemblies in the storage pool and requires the associated
crane interlocks and physical stops be periodically demonstrated
operable. During installation, racks and associated handling tools
will be moved over the spent fuel pool, however there will be no
fuel in the pool when the 24 flux trap rack modules are installed. A
three foot lateral free zone clearance from stored spent fuel will
be maintained during installation of the ten smaller burnup credit
rack modules. Installation work in the spent fuel pit area will be
controlled and performed in strict accordance with specific written
instructions.
NUREG-0612 states that in lieu of providing a single failure-
proof crane system, the control-of-heavy-loads guidelines can be
satisfied by establishing that the potential for a heavy load drop
is extremely small. Storage rack movements to be accomplished with
the WBN Auxiliary Building crane will conform with NUREG-0612
guidelines in that the probability of a drop of a storage rack is
extremely small. The crane has a tested capacity of 125 tons. The
maximum weight of any existing, replacement, or new storage rack and
its associated handling tool is less than 20 tons. Therefore, there
is ample safety factor margin for movements of the storage racks by
the Auxiliary Building crane. Special lifting devices, which have
redundancy or a rated capacity sufficient to maintain adequate
safety factors, will also be utilized in the movements of the
storage racks. In accordance with NUREG-0612, Appendix B, the safety
margin ensures that the probability of a load drop is extremely low.
Future load travel over fuel stored in a rack specifically
designed for the cask loading area of the cask pit will be
prohibited unless an impact shield, which has been specifically
designed for this purpose, is covering the area. Loads that are
permitted when the shield is in place must meet analytically
determined weight, travel height, and cross-sectional area criteria
that preclude penetration of the shield. A Technical Requirement
(TR) has been proposed that incorporates the previously mentioned
load criteria.
Also a rack change-out sequence is being developed that
addresses removal of the existing racks, movement of the new racks
into the Auxiliary Building, initial staging on the refueling floor,
and final installation in the pool. The change-out sequence
objectives include establishing lift heights, travel distances, and
number of lifts to be as low as reasonably achievable. Accordingly,
it is concluded that the proposed installation activities will not
significantly increase the probability of a load-handling accident.
The consequences of a load-handling accident are unaffected by the
proposed installation activities.
The consequences of a spent fuel assembly drop were evaluated,
and it was determined that the racks will not be distorted such that
the racks would not perform their safety function. The criticality
acceptance criterion, Keff less than or equal to 0.95, is not
violated, and the calculated doses are well within 10 CFR Part 100
guidelines. The radiological consequences of the fuel assembly drop
accident evaluated for WBN, have changed, however, the changes do
not involve a significant increase in consequences and are well
within the 10 CFR 100 requirements.
A TRM change has been proposed that would permit the transfer-
canal gate and the divider gate for the cask pit to travel over fuel
assemblies in the spent fuel pool during movement between their
gated and stored position. Rack damage is restricted to an area
above the active fuel region, therefore, neither criticality nor
radiological concerns exist.
The consequences of a seismic event have been evaluated. The
replacement racks are
[[Page 15735]]
designed and fabricated and the new racks will be fabricated to meet
the requirements of applicable portions of the NRC regulatory guides
and published standards. Design margins have been provided for rack
tilting, deflection, and movement such that the racks do not impact
each other or the spent fuel pool walls in the active fuel region
during the postulated seismic events. The free-standing racks will
maintain their integrity during and after a seismic event. The fuel
assemblies also remain intact and therefore no criticality concerns
exist.
The spent fuel pool system is a passive system with the
exception of the fuel pool cooling train and heating, ventilating,
and air-conditioning (HVAC) equipment. Redundancies in the cooling
train and HVAC hardware are not reduced by the planned fuel storage
modification. The potential increased heat load resulting from any
additional storage of spent fuel is well within the existing system
cooling capacity. Therefore, the probability of occurrence or
malfunction of safety equipment leading to the loss-of-cooling flow
in the spent fuel pool is not significantly affected. Furthermore,
the consequences of this type incident are not significantly
increased from previously evaluated cooling system loss of flow
malfunctions. Thermal-hydraulic scenarios assume the reracked pool
is approximately 90% full with spent fuel assemblies. From this
starting point, the remaining storage capacity is utilized by
analyzing both normal and unplanned full core off loads using
conservative assumptions and previously established methods.
Calculated values include maximum pool water bulk temperature,
coincident maximum pool water local temperature, the maximum fuel
cladding temperature, time-to-boil after loss-of-cooling paths, and
the effect of flow blockage in a storage cell.
Although the proposed modification increases the pool heat load,
results from the above analyses yield a maximum bulk temperature
less than 160 degrees Fahrenheit which is below the bulk boiling
temperature. Also the maximum local water temperature is below
nucleate boiling condition values. Associated results from
corresponding loss-of-cooling evaluations give minimums of 5.3 hours
before boiling begins and 45 hours before the pool water level drops
to the minimum required for shielding spent fuel.
This is sufficient time to begin utilization of available
alternate sources of makeup cooling water. Also, the effect of the
increased thermal loading on the pool structure, associated cooling
system, and components was evaluated and determined to establish an
acceptable design basis with the new storage configuration. No
modifications were necessary because of the increased temperature.
(2) Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously analyzed.
The proposed modification has been evaluated in accordance with
the guidance of the NRC position paper entitled, ``OT Position for
Review and Acceptance of Spent-Fuel Storage and Handling
Applications'', appropriate NRC regulatory guidelines; appropriate
NRC standard review plans; and appropriate industry codes and
standards. Proven analytical technology was used in designing the
planned fuel storage expansion and will be utilized in the
installation process. Basic reracking technology has been developed
and demonstrated in applications for fuel pool capacity increases
that have already received NRC staff approval.
Proposed TSs for the spent fuel storage racks use burnup credit
and fuel assembly administrative placement restrictions for
criticality control. These restrictions are described in the
proposed change to the design features section of the TSs by
reference to the Spent Fuel Pool Modifications report. Additional
evaluations were required to ensure that the criticality criterion,
keff less than or equal to 0.95, is maintained. These include
evaluation for the abnormal placement of unirradiated (fresh) fuel
assemblies of 5.0 wt% enrichment into a storage cell location
designed for lower enrichment or irradiated fuel. Soluble boron, for
which credit is permitted under these abnormal conditions, ensures
that reactivity is maintained substantially less than the design
requirement. For example, if the PaR flux trap racks are
inadvertently all loaded with fresh assemblies of the maximum 5.0
wt% fuel instead of observing the 3.8 wt% and 6.75 MWD/KgU controls,
the worth of the 2000 ppm borated water is sufficient to lower the
keff of the storage racks to 0.83. The existing and proposed
TSs require boron concentration in the pool and cask pit to be
greater than or equal to 2000 ppm during fuel movement. An
analytical determination of the reactivity worth of 2000 ppm borated
water in the spent fuel storage pool predicted the change in
keff to be approximately 17 percent keff. Although no
credit for soluble boron was proposed in the TSs, it was also
determined by an independent calculation that a minimum
concentration of 520 ppm soluble boron allows the unrestricted
storage of 5.0 wt% enriched fuel in the PaR flux trap racks.
The Holtec-designed peripheral ``baby'' racks and the 15 x 15
racks in the cask loading area can safely and conservatively store
fuel of 5 wt% initial enrichment burned to 41 MWD/kgU or lower
enriched fuel with lower burnup, i.e., fuel of equivalent
reactivity. Evaluations have confirmed that, for the abnormal
placement of a fresh fuel assembly of 5.0 wt% in these racks, the
criticality criterion is maintained with the existing and proposed
TS requirements of 2000 ppm soluble boron.
Although these changes required addressing additional aspects of
a previously analyzed accident, the possibility of a previously
unanalyzed accident is not created.
The impact shield design together with its attendant
administrative controls and NUREG-0612 heavy load lift compliance,
renders the possibility of a heavy load drop on fuel as not credible
in accordance with the NUREG-0612 single-failure-proof criteria.
Accordingly, since this particular part of the proposed reracking
modification is not a change that could malfunction by a new single
failure, the movement of heavy loads over the cask pit does not
create the possibility of a new or different kind of accident.
It is therefore concluded that the proposed reracking does not
create the possibility of a new or different kind of accident from
any previously analyzed.
(3) Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
The design and technical review process applied to the reracking
modification included addressing the following areas:
1. Nuclear criticality considerations.
2. Thermal-hydraulic considerations.
3. Mechanical, material, and structural considerations.
The established acceptance criterion for criticality is that the
neutron multiplication factor shall be less than or equal to 0.95,
including all uncertainties. The results of the criticality analyses
for the rack designs demonstrate that this criterion is satisfied.
The methods used in the criticality analysis conform to the
applicable portions of NRC guidance and industry codes, standards,
and specifications. In meeting the acceptance criteria for
criticality in the spent fuel pool and the cask loading area, such
that keff is always less than 0.95 at a 95/95 percent
probability tolerance level, the proposed amendment does not involve
a significant reduction in the margin of safety for nuclear
criticality.
Conservative methods and assumptions were used to calculate the
maximum fuel temperature and the increase in temperature of the
water in the spent fuel pit area. The thermal-hydraulic evaluation
used methods previously employed. The proposed storage modification
will increase the heat load in the spent fuel pool, but the
evaluation shows that the existing spent fuel cooling system will
maintain the bulk pool water temperature at or below 160 degrees
Fahrenheit. Thus it is demonstrated that the worst-case peak value
of the pool bulk temperature is considerably lower than the bulk
boiling temperature. Evaluation also shows that maximum local water
temperatures along the hottest fuel assembly are below the nucleate
boiling condition value. Thus, there is no significant reduction in
the margin of safety for thermal hydraulic or spent fuel cooling
considerations.
The mechanical, material, and structural design of the spent
fuel racks is in accordance with applicable portions of NRC's
position in ``OT Position for Review and Acceptance of Spent-Fuel
Storage and Handling Applications,'' dated April 14, 1978 (as
modified January 18, 1979), as well as other applicable NRC guidance
and industry codes. The primary safety function of the spent fuel
racks is to maintain the fuel assemblies in a safe configuration
through normal and abnormal loading conditions. Abnormal loadings
that have been evaluated with acceptable results and discussed
previously include the effect of an earthquake and the impact
because of the drop of a fuel assembly. The rack materials used are
compatible with the fuel assemblies and the environment in the spent
fuel pool.
[[Page 15736]]
The structural design for the new racks provides tilting,
deflection, and movement margins such that the racks do not impact
each other or the spent fuel pit walls in the active fuel region
during the postulated seismic events. Also the spent fuel assemblies
themselves remain intact and no criticality concerns exist. In
addition, finite element analysis methods were used to evaluate the
continued structural acceptability of the spent fuel pit. The
analysis was performed in accordance with ``Building Code
Requirements for Reinforced Concrete,'' (ACI 318-63,77). Therefore,
with respect to mechanical, material, and structural considerations,
there is no significant reduction in a margin of safety.
Summary
Based on the above analysis, TVA has determined that operation
of WBN, in accordance with the proposed amendment, would not: (1)
involve a significant increase in the probability of consequences of
an accident previously evaluated, (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. Therefore, operations of WBN in accordance with the proposed
amendments as described do not involve significant hazard
considerations as defined in 10 CFR 50.92 and that the criteria of
10 CFR 50.91 have accordingly been met.
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 thirty (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 20555-0001, and should cite the publication date and
page number of this Federal Register notice. Written comments may also
be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the NRC Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC.
The filing of requests for hearing and petitions for leave to
intervene is discussed below.
By May 2, 1997, the licensee may file a request for a hearing with
respect to issuance of the amendment to the subject facility operating
license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room located at the Chattanooga-Hamilton County Library, 1001
Broad Street, Chattanooga, Tennessee 37402. 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 a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing.
The petitioner must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner intends to rely to establish those facts or expert opinion.
Petitioner must provide sufficient information to show that a genuine
dispute exists with the applicant on a material issue of law or fact.
Contentions shall be limited to matters within the scope of the
amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner to relief. A petitioner who fails
to file such a supplement which satisfies these requirements with
respect to at least one contention will not be permitted to participate
as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the 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,
[[Page 15737]]
notwithstanding the request for a hearing. Any hearing held would take
place after issuance of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Docketing and
Services Branch, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC,
by the above date. 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 Mr. Frederick J. Hebdon: 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, and to
General Counsel, Tennessee Valley Authority, ET 10H, 400 West Summit
Hill Drive, Knoxville, Tennessee 37902, 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 41670, 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 may grant an untimely request for oral argument only
upon showing of good cause by the requesting party 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 the 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, October 23, 1996, as supplemented on
December 11, 1996, January 31, February 10 and 24 and March 11, 1997
which is available for public inspection at the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC,
and at the local public document room, located at the Chattanooga-
Hamilton County Library, 1001 Broad Street, Chattanooga, Tennessee.
Dated at Rockville, Maryland, this 27th day of March 1997.
For the Nuclear Regulatory Commission.
Robert E. Martin,
Sr. Project Manager, Project Directorate II-3, Division of Reactor
Projects--I/II, Office of Nuclear Reactor Regulation.
[FR Doc. 97-8401 Filed 4-1-97; 8:45 am]
BILLING CODE 7590-01-P
