Niagara Mohawk Power Corporation; Notice of Consideration of Issuance of Amendment to Facility Operating License, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing

[Federal Register Volume 63, Number 226 (Tuesday, November 24, 1998)]
[Notices]
[Pages 64973-64976]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-31336]


=======================================================================
-----------------------------------------------------------------------

NUCLEAR REGULATORY COMMISSION

[Docket No. 50-220]


Niagara Mohawk Power Corporation; 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 or NRC) is 
considering issuance of an amendment to Facility Operating License No. 
DRP-63 issued to Niagara Mohawk Power Corporation (NMPC or the 
licensee) for operation of Nine Mile Point Nuclear Station, Unit 1 
(NMP1), located in the town of Scriba, Oswego County, New York.
    The proposed amendment would change Technical Specification (TS) 
5.5, ``Storage of Unirradiated and Spent Fuel,'' for NMP1. The changes 
would reflect a planned modification to increase the number of fuel 
assemblies that can be stored in the spent fuel pool from 2776 to 4086. 
The changes would also delete an erroneous reference within TS 5.5 to 
10 CFR 70.55 for calculational methods approved by the Commission 
involving special arrays.
    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 operation of NMP1, in accordance with the proposed 
amendment, will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    Analysis of issues concerning the expanded spent fuel pool 
storage capacity modification has considered the following potential 
scenarios:
    1. A spent fuel assembly drop in the spent fuel pool.
    2. Loss of spent fuel pool cooling flow.
    3. A seismic event.
    4. A cask drop in the spent fuel pool.
    5. An accidental drop of a rack module during construction 
activity in the pool.
    The probability that any of the first four scenarios in the 
above list can occur is not significantly increased by the proposed 
Technical Specification changes and the associated modification 
activities. Spent fuel pool activities such as fuel assembly 
movement as well as Spent Fuel Pool Cooling System operation will 
continue to be performed in accordance with approved plant 
procedures. A cask drop into the pool is considered an unlikely 
event based on the design/maintenance of the main hoist, the 
controlled cask movement path and the cask drop protection system 
(hydraulic guide cylinder). None of these features are affected by 
the proposed change. Concerning installation activities, whether 
conducted during power operation or shutdown, the reactor building 
crane will be utilized for handling all heavy loads (i.e., old and 
new racks) during the reracking operation. The main hoist is 
equipped with a redundant hoisting system which will prevent the 
dropping of heavy loads in the event that a cable or other critical 
part of the main hoist equipment should fail. Operability of the 
cranes will be checked and verified before the re-racking operation. 
All lift rigging and the refueling crane/hoist system will be 
inspected and all heavy load lifts will comply with NUREG-0612, 
``Control of Heavy Loads at Nuclear Power Plants,'' per plant 
procedures. Accordingly, the probability of a heavy load drop will 
not significantly increase.
    Therefore, the proposed modification and associated Technical 
Specification changes do not involve a significant increase in the 
probability of an accident previously evaluated.
    UFSAR [Updated Final Safety Analysis Report] Section 15.c.3, 
``Refueling Accident,'' discusses the accident in which a fuel 
bundle is accidently dropped onto the top of the core during 
refueling operations and the subsequent radiological effects. Fuel 
assembly density in the core is essentially equivalent to that of 
the assemblies stored in the replacement spent fuel racks. 
Accordingly, the consequence of a fuel assembly dropped on the core 
(as analyzed in UFSAR Section 15.c.3), is not significantly

[[Page 64974]]

increased. Also, analysis shows that such an accident will not 
distort the racks sufficiently to impair their functionality and the 
minimum subcriticality margin, keff [neutron 
multiplication factor] [less than or equal to] 0.95, will be 
maintained. Thus, the consequences of such an accident remain 
acceptable and are not greater than those of previously evaluated 
accidents.
    The consequences of a loss of spent fuel pool cooling have been 
evaluated and found acceptable. In the unlikely event that all 
pooling cooling is lost, sufficient time is available for the 
operators to re-establish cooling before the onset of pool boiling. 
Also, the consequences of a design basis seismic event have been 
evaluated and found acceptable. The new and the existing racks have 
been analyzed in their new configuration and found safe and impact-
free during seismic motion. The structural capability of the pool 
will not be exceeded under dead weight, thermal, and seismic loads 
and the reactor building and the crane structure will retain the 
necessary safety margins during a seismic event. Thus, the 
consequences of a seismic event are not significantly increased.
    Movements of heavy loads over the pool will continue to comply 
with applicable guidelines (e.g., NUREG-0612) and procedures. As 
previously mentioned, no heavy loads (e.g., racks, casks) will be 
transported over any region of the spent fuel pool containing fuel. 
The consequences of an accidental drop of a rack module into the 
pool during reracking activities have been evaluated indicating that 
very limited damage to the liner could occur. Therefore, the 
consequences of a heavy load drop are not increased.
    During rack removal and installation activities, interim 
configurations will exist (i.e., various combinations of old and new 
racks). These combinations have been evaluated and indicate that no 
thermal-hydraulic, criticality and structural concerns exist.
    The last paragraph in Section 5.5 states that calculations for 
keff values have been based on methods approved by the 
NRC covering special arrays (10 CFR 70.55). 10 CFR 70.55, 
``Inspections,'' discusses inspections of special nuclear material 
and the premises and facilities where special nuclear material is 
used; not methods used to determine keff. Therefore, this 
is an inaccurate reference. Also, although the NRC does review and 
approve our methods to determine keff (as part [of] the 
Technical Specification Amendment approval process) this information 
is not considered critical design feature information. Accordingly, 
it does not belong in Section 5.0, ``Design Features,'' of the 
Technical Specifications. Based on the above, deletion of this 
paragraph will not have any adverse affect on safety and will 
eliminate any potential confusion involving the reference to 10 CFR 
70.55.
    Therefore, the proposed changes do not significantly increase 
the consequences of any accident previously evaluated.
    The operation of NMP1, in accordance with the proposed 
amendment, will not create the possibility of a new or different 
kind of accident from any accident previously evaluated. 
    The proposed modification activities and associated Technical 
Specification amendment does not introduce any new modes of plant 
operation or accident precursors which could initiate a new or 
different kind of accident, affect the operation or function of any 
equipment necessary for the safe operation or shutdown of the plant, 
or involve any changes to plant operating parameters. The only 
physical alterations of plant configuration will involve the removal 
of currently installed non-poison and Boraflex spent fuel racks and 
the installation of new high density Boral racks. Heavy load 
movements (i.e., the old and new racks, casks) will continue to be 
performed in accordance with NUREG-0612. Accordingly, a drop of 
heavy loads onto spent fuel during and following installation 
activities need not be considered. As previously discussed, 
installation of the new racks does not constitute a thermal-
hydraulic, criticality or structural concern. Therefore, this change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The operation of NMP1, in accordance with the proposed 
amendment, will not involve a significant reduction in a margin of 
safety. 
    The proposed modification activities and associated Technical 
Specification Amendment involves replacing the currently install 
non-poison flux trap and Boraflex storage racks with new high 
density Boral racks. The proposed Technical Specification changes 
will not reduce the equipment required by Technical Specifications, 
affect any Technical Specification system setpoints, or adversely 
affect the ability of plant equipment to respond to an accident.
    The design and technical considerations applied to the reracking 
modification included addressing the following areas:
    1. Nuclear criticality considerations.
    2. Thermal-hydraulic considerations.
    3. Mechanical, material and structural considerations.
    Concerning criticality considerations, the replacement high 
density spent fuel storage racks are designed to assure that the 
neutron multiplication factor ( keff ) is equal to or 
less than 0.95 with the racks fully loaded with fuel of the highest 
anticipated reactivity and the pool flooded with unborated water at 
a temperature corresponding to the highest reactivity. The maximum 
calculated reactivity includes a margin for uncertainty in 
reactivity calculations and in mechanical tolerances, statistically 
combined, such that the true keff will be equal to or 
less than 0.95 with a 95% probability at a 95% confidence level. 
Reactivity effects of abnormal and accident conditions have also 
been evaluated to assure that under credible abnormal conditions, 
the reactivity will be less than the limiting design basis value. 
Accordingly, the proposed change does not involve a significant 
reduction in a margin of safety in that the existing racks maintain 
a keff of less than 0.95.
    Amendment No. 54 to the NMP1 [Operating License which changed 
the] Technical Specifications, dated February 1, 1984, increased the 
spent fuel storage capacity to the current maximum of 2776 
assemblies. In [its] Safety Evaluation, Section 2.4, ``Spent Fuel 
Pool Cooling Considerations,'' the NRC indicated acceptance of 
NMPC's thermal-hydraulic analysis based on: (1) with the maximum 
normal heat load assumed and one cooling train in operation, pool 
water is calculated to 125 degrees F which is below the 140 degrees 
F limit recommended in Standard Review Plan (SRP) Section 9.1.3; and 
(2) with the maximum abnormal heat load assumed and two cooling 
trains operating, the maximum pool temperature is calculated to be 
below 124 degrees which is below the boiling temperature limit set 
forth in SRP Section 9.1.3.
    The SRP requires that with a maximum normal heat load and a 
single failure, pool temperatures should be kept below 140 degrees F 
and that with an abnormal heat load, pool temperatures should be 
kept below boiling. For the abnormal heat load case, consideration 
of a single failure is not required. The analysis provided in 
Section 5, Attachment C of this submittal [the licensee's May 15, 
1998] indicates how the proposed change meets the requirements of 
the SRP and, accordingly, that no significant decrease in a margin 
of safety occurs.
    In SRP 9.1.3, a normal spent fuel pool heat load is considered 
to be a core shuffle. NMPC has evaluated the core shuffle using the 
SRP guidance as Case 1, in previously referenced Section 5 of 
Attachment C. This evaluation indicates that a maximum pool 
temperature of 119 degrees F will be reached, thereby meeting the 
SRP maximum temperature requirement of 140 degrees F. Because a 
``normal heat load'' now potentially involves a full core offload, 
NMPC has also reviewed this discharge scenario (Case 3, Section 5) 
as a normal case and therefore assumed a single failure. As 
delineated in Case 3, calculations will be performed to determine 
the days after reactor shutdown when all assemblies can be 
transferred to the pool, as a function of reactor building cooling 
water temperatures, such that a 140 degrees F bulk pool temperature 
will not be exceeded. Therefore, the SRP bulk pool temperature limit 
of 140 degrees F for a maximum normal heat load (both shuffle and 
full core offload) will not be exceeded.
    The SRP also requires that for an abnormal maximum heat load 
(emergency condition), without a single failure, that pool 
temperatures should be maintained below boiling. Using the 
guidelines provided in the SRP, calculations were performed that 
found the maximum pool temperature to be 135 degrees F which is well 
below the SRP criteria (Case 2).
    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

[[Page 64975]]

normal and abnormal loading conditions. Abnormal loadings that have 
been evaluated with acceptable results include the effect of an 
earthquake and the impact due to the drop of a fuel assembly. The 
rack materials used are compatible with the fuel assemblies and the 
environment in the spent fuel pool. 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 pool 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, the structural adequacy of 
the spent fuel pool was demonstrated.
    During rack removal and installation activities, interim 
configurations will exist (i.e., various combinations of old and new 
racks). These combinations have been evaluated and indicate that no 
thermal-hydraulic, criticality and structural concerns exist.
    Therefore, the proposed change will not result in a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based upon 
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.
    By December 24, 1998, 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 such 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 Reference and Documents Department, 
Penfield Library, State University of New York, Oswego, New York 13126. 
If a request for a hearing and 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 
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.
    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.
    A request for a hearing and 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: Rulemakings and 
Adjudications Staff, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC, 
by the above date. A copy of the petition should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and to Mr. Mark J. Wetterhahn, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502, attorney for the 
licensee.

[[Page 64976]]

    Untimely 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)(l)-(v) and 2.714(d).
    Pursuant to the Commission's regulations, 10 CFR 2.1107, 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 Fuel Storage Capacity at Civilian Nuclear Power 
Reactors'' (published at 50 FR 41662 dated October 15, 1985). 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 ten (10) days of an order granting a request for hearing 
or petition to intervene. The presiding officer must grant a timely 
request for oral argument. The presiding officer may grant an untimely 
request for oral argument only upon a 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 must 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 an adjudicatory 
hearing. If no party to the proceeding timely requests oral argument, 
and 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 May 15, 1998, as supplemented September 
25 and October 13, 1998, which are available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room 
located at the Reference and Documents Department, Penfield Library, 
State University of New York, Oswego, New York 13126.

    Dated at Rockville, Maryland, this 18th day of November 1998.

    For the Nuclear Regulatory Commission.
Darl S. Hood,
Senior Project Manager, Project Directorate I-1, Division of Reactor 
Projects--I/II, Office of Nuclear Reactor Regulation.
[FR Doc. 98-31336 Filed 11-23-98; 8:45 am]
BILLING CODE 7590-01-P