[Federal Register Volume 62, Number 83 (Wednesday, April 30, 1997)]
[Notices]
[Pages 23499-23501]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-11120]
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NUCLEAR REGULATORY COMMISSION
[Docket No. 50-254 and 50-265]
Commonwealth Edison company and Midamerican Energy Company;
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The U.S. Nuclear Regulatory Commission (the Commission) is
considering issuance of amendments to Facility Operating License Nos.
DPR-29 and DPR-30, issued to Commonwealth Edison Company (ComEd, the
licensee), for operation of the Quad Cities Nuclear Power Station,
Units 1 and 2, located in Rock Island County, Illinois.
The proposed amendments would reflect a change in the Quad Cities,
Unit 2, Minimum Critical Power Ratio (MCPR) Safety Limit and add the
Siemens Power Corporation (SPC) methodology for application of the
Advanced Nuclear Fuel for Boiling Water Reactors (ANFB) Critical Power
Correlation to coresident General Electric fuel for Quad Cities, Unit
2, Cycle 15, to Technical Specification (TS) Section 6.9.A.6.b.
This request for amendments was submitted under exigent
circumstances to support Quad Cities, Unit 2, Cycle 15, operation which
is scheduled to be on line May 19, 1997. On March 20, 1997, SPC
determined the need for a larger data base for determining the additive
constant uncertainty. The combined time necessary for SPC to develop
the new data base and the time for ComEd to develop this TS request
would not allow the normal 30-day period for public comment to support
Quad Cities, Unit 2, startup.
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.
Pursuant to 10 CFR 50.91(a)(6), for amendments to be granted under
exigent circumstances, the NRC staff must determine that the requested
amendments involve 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 amendments 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. Involve a significant increase in the probability or
consequences of an accident previously evaluated:
The probability of an evaluated accident is derived from the
probabilities of the individual precursors to that accident. The
consequences of an evaluated accident are determined by the
operability of plant systems designed to mitigate those
consequences. Limits have been established consistent with NRC
approved methods to ensure that fuel performance during normal,
transient, and accident conditions is acceptable. The proposed
Technical Specifications amendment conservatively establishes the
MCPR Safety Limit for Quad Cities Unit 2, such that the fuel is
protected during normal operation and during any plant transients or
anticipated operational occurrences. Additionally, methodologies are
being added to the Section 6.9.A.6.b list of methodologies utilized
in determining core operating limits.
a. MCPR Safety Limit and MCPR Safety Limit Bases Change
The probability of an evaluated accident is not increased by
increasing the MCPR Safety Limit to 1.10 and changing the MCPR
Safety Limit Bases. The change does not require any physical plant
modifications, physically affect any plant components, or entail
changes in plant operation. Therefore, no individual precursors of
an accident are affected.
This Technical Specification amendment proposes to change the
MCPR Safety Limit to protect the fuel during normal operation as
well as during any transients or anticipated operational
occurrences. The method that is used to determine the ATRIUM-9B
additive constant uncertainty is conservative, such that, the
resulting MCPR Safety Limit is high enough to ensure that less than
0.1% of the fuel rods are expected to experience boiling transition
if the limit is not violated. Operational limits will be established
based on the proposed MCPR Safety Limit to ensure that the MCPR
Safety Limit is not violated during all modes of operation. This
will ensure that the fuel design safety criteria, more than 99.9% of
the fuel rods avoiding transition boiling during normal operation as
well as anticipated operational occurrences, is met. The method for
calculating an ATRIUM-9B additive constant uncertainty, is described
in Reference 2 [SPC document, ANFB Critical Power Correlation
Uncertainty For Limited Data Sets, ANF-1125(P), Supplement 1,
Appendix D, Siemens Power Corporation--Nuclear Division, Submitted
on April 18, 1997] and is based on an expanded pool of data for the
ATRIUM-9B fuel design (527 data points). The additive constant
uncertainty from Reference 2 is then used to determine the change
from the additive constant uncertainty using the original pool of
data (125 data points). This difference is conservatively doubled
and added to the additive constant uncertainty using the original
pool of data (125 data points). Reference 5 [Siemens Power
Corporation letter, ``Interim Use of Increased ANFB Additive
Constant Uncertainty'', HDC:97:033, H.D. Curet to Document Control
Desk, April 18, 1997] documents the conservative interim approach of
doubling the difference in additive constant uncertainties. The
resulting additive constant uncertainty is used to determine the
Quad Cities Unit 2 Cycle 15 MCPR Safety Limit. Since the new MCPR
Safety Limit was determined using a conservative ATRIUM-9B additive
constant uncertainty, and the operability of plant systems designed
to mitigate any consequences of accidents have not changed, the
consequences of an accident previously evaluated are not expected to
increase.
b. Addition of Siemens Power Corporation's (SPC) methodology for
Application of the ANFB Critical Power Correlation to Coresident GE
Fuel for Quad Cities Unit 2 Cycle 15 to Section 6.9.A.6.b
[[Page 23500]]
The probability of an evaluated accident is not increased by
adding Reference 1 [ComEd letter, ``ComEd Response to NRC Staff
Request for Additional Information (RAI) Regarding the Application
of Siemens Power Corporation ANFB Critical Power Correlation to
Coresident General Electric Fuel for LaSalle Unit 2 Cycle 8 and Quad
Cities Unit 2 Cycle 15, NRC Docket No.'s 50-373/374 and 50-254/
265'', J.B. Hosmer to U.S. NRC, July 2, 1996, transmitting the
topical report, Application of the ANFB Critical Power Correlation
to Coresident GE Fuel for Quad Cities Unit 2 Cycle 15, EMF-96-
051(P), Siemens Power Corporation--Nuclear Division, May 1996, and
related information], to Section 6.9.A.6.b. Reference 1 describes
the methodology used to determine the additive constants and the
associated uncertainty of the Quad Cities Unit 2 Cycle 15 GE9 and
GE10 fuel for the ANFB critical power correlation. The additive
constant and the associated uncertainties for the GE9 and GE10 fuel
are used to calculate the MCPR Safety Limit, which in turn is used
to establish the MCPR operating limit for Quad Cities Unit 2 Cycle
15 operation. Therefore, adding Reference 1 to Section 6.9.A.6.b of
the Technical Specifications updates the Reference list to include a
methodology used for determining Quad Cities Unit 2 Cycle 15
operational limits.
Adding Reference 1 to the Reference list in Section 6.9.A.6.b
also will not increase the consequences of an accident previously
evaluated. Reference 1 determines the additive constants and the
associated uncertainty for the GE fuel in Quad Cities Unit 2 Cycle
15. It also provides input for determining the MCPR Safety Limit.
Because Reference 1 contains conservative methods and calculations
and because the operability of plant systems designed to mitigate
any consequences of accidents have not changed, the consequences of
an accident previously evaluated will not increase.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated:
Creation of the possibility of a new or different kind of
accident would require the creation of one or more new precursors of
that accident. New accident precursors may be created by
modifications of the plant configuration, including changes in
allowable modes of operation. This Technical Specification submittal
does not involve any modifications of the plant configuration or
allowable modes of operation. This Technical Specification submittal
involves a) an added conservatism in the Quad Cities Unit 2 MCPR
Safety Limit due to analytical changes and use of an expanded
database, and b) an additional reference incorporated in Section
6.9.A.6.b describing the methodology used to determine the additive
constants and additive constant uncertainty for GE9 and GE10 fuel
for Quad Cities Unit 2 Cycle 15. Therefore, no new precursors of an
accident are created and no new or different kinds of accidents are
created.
3. Involve a significant reduction in the margin of safety for
the following reasons:
The MCPR Safety Limit provides a margin of safety by ensuring
that less than 0.1% of the rods are expected to be in boiling
transition if the MCPR limit is not violated. The proposed Technical
Specification amendment reflects MCPR Safety Limit results from
conservative calculations by SPC using the new ATRIUM-9B additive
constant uncertainty. These new ATRIUM-9B additive constant
uncertainty calculations are based on a larger pool of data than
previous calculations (527 data points versus 125 data points).
Additionally, the additive constant uncertainty resulting from
statistical analyses of the larger pool of data is conservatively
applied to calculate a new MCPR Safety Limit of 1.10, which is more
restrictive than the current MCPR Safety Limit of 1.07.
SPC has increased its ATRIUM-9B critical power test data base
from 125 data points at 1000 psi with mass fluxes ranging from 0.5
to 1.5 Mlb/hr-ft2, to 527 data points that cover a wider
range of operating pressures, flows, and axial power shapes.
The Experimental Critical Power Ration (ECPR) and the standard
deviation of the ECPR for each of the 527 data points are
statistically examined by an Analysis of Variance. The results of
the Analysis of Variance of the Pressure Groups are a mean ECPR, a
standard deviation of ECPR, degrees of freedom, and equivalent
sample size.
The overall uncertainty for CPR is statistically calculated
using the standard deviation of the pooled data and the variance
between the means associated with the axial power shapes. An upper
95% confidence limit standard deviation is calculated based on Chi-
Square for the calculated degrees of freedom. This overall standard
deviation in ECPR is converted to an additive constant uncertainty.
This conversion is derived from the ratios of the ANFB correlation
standard deviation to the additive constant standard deviation for
the ATRIUM-9B data.
This calculated additive constant uncertainty is not directly
applied to the MCPR Safety Limit calculation. A conservative ATRIUM-
9B additive constant uncertainty is used to calculate a new MCPR
Safety Limit for Quad Cities Unit 2 Cycle 15.
The difference is calculated between the additive constant
uncertainties after and prior to the data set being expanded to
include 527 points. This difference is then conservatively doubled
and added to the additive constant uncertainty prior to the
expansion of the data set (based on 125 data points).
The resulting additive constant uncertainty, 0.029, is used to
calculate a new MCPR Safety Limit value of 1.10 for Quad Cities Unit
2 Cycle 15.
Because a conservative method is used to apply the ATRIUM-9B
additive constant uncertainty to the MCPR Safety Limit calculation,
a decrease in the margin of safety will not occur due to changing
the MCPR Safety Limit. The revised Safety Limit will ensure the
appropriate level of fuel protection. Additionally, operational
limits will be established based on the proposed MCPR Safety Limit
to ensure that the MCPR Safety Limit is not violated during all
modes of operation. This will ensure that the fuel design safety
criteria, more than 99.9% of the fuel rods avoiding transition
boiling during normal operation as well as anticipated operational
occurrences, is met.
The margin of safety is not decreased by adding the Reference to
Section 6.9.A.6.b of Siemens Power Corporation's (SPC) methodology
for application of the ANFB Critical Power Correlation to coresident
GE Fuel for Quad Cities Unit 2 Cycle 15. While this methodology is
in review by the NRC, and pending approval for application to Quad
Cities Unit 2 Cycle 15, it is the same methodology previously
reviewed and approved for use at LaSalle Unit 2 (References 3 and 4)
[ComEd letter, ``Application of Siemen's Power Corporation ANFB
Critical Power Correlation to Coresident General Electric Fuel for
LaSalle Unit 2 Cycle 8'', G.G, Benes to U.S. Nuclear Regulatory
Commission, dated March 8, 1996, and NRC SER letter, ``Safety
Evaluation for Topical Report EMF-96-021(P), Revision 1,
`Application of the ANFB Critical Power Correlation to Coresident GE
Fuel for LaSalle Unit 2 Cycle 8' (TAC No. M94964)'', D.M. Skay to I.
Johnson, dated September 26, 1996.
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
requested amendments involve no significant hazards consideration.
The Commission is seeking public comments on this proposed
determination. Any comments received within 14 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendments until the
expiration of the 14-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 amendments before the expiration
of the 14-day notice period, provided that its final determination is
that the amendments involve 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. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, 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,
[[Page 23501]]
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 30, 1997, the licensee may file a request for a hearing with
respect to issuance of the amendments 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 Dixon Public Library, 221 Hennepin Avenue,
Dixon, Illinois 61021. 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 amendments 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 the amendments are issued before the expiration of the 30-day
hearing period, the Commission will make a final determination on the
issue of no significant hazards consideration. If a hearing is
requested, the final determination will serve to decide when the
hearing is held.
If the final determination is that the amendments requested involve
no significant hazards consideration, the Commission may issue the
amendments and make them immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendments.
If the final determination is that the amendments requested involve
a significant hazards consideration, any hearing held would take place
before the issuance of any amendments.
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 Robert A. Capra: 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 20555-0001, and to
Michael I. Miller, Esquire; Sidley and Austin, One First National
Plaza, Chicago, Illinois 60603, 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 amendments dated April 21, 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 Dixon Public Library, 221 Hennepin
Avenue, Dixon, Illinois 61021.
Dated at Rockville, Maryland, this 24th day of April 1997.
For the Nuclear Regulatory Commission.
Robert M. Pulsifer,
Project Manager, Project Directorate III-2, Division of Reactor
Projects--III/IV, Office of Nuclear Reactor Regulation.
[FR Doc. 97-11120 Filed 4-29-97; 8:45 am]
BILLING CODE 7590-01-P