[Federal Register Volume 71, Number 7 (Wednesday, January 11, 2006)]
[Notices]
[Pages 1774-1776]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E6-159]
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NUCLEAR REGULATORY COMMISSION
[Docket No. 50-271]
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear
Operations, Inc.; Notice of Consideration of Issuance of Amendment to
Facility Operating License and Proposed No Significant Hazards
Consideration Determination
The U.S. Nuclear Regulatory Commission (NRC or the Commission) is
considering issuance of an amendment to Facility Operating License No.
DPR-28, issued to Entergy Nuclear Vermont Yankee, LLC and Entergy
Nuclear Operations, Inc. (the licensee), for operation of the Vermont
Yankee Nuclear Power Station (VYNPS) located in Windham County,
Vermont.
The proposed amendment would change the VYNPS operating license to
increase the maximum authorized power level from 1593 megawatts thermal
(MWt) to 1912 MWt. This change represents an increase of approximately
20 percent above the current maximum authorized power level. The
proposed extended power uprate (EPU) amendment would also change the
VYNPS Technical Specifications (TSs) to provide for implementing
uprated power operation.
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 Title 10 of the Code of Federal Regulations
(10 CFR), Sec. 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. The NRC
staff's analysis of the issue of no significant hazards consideration
is presented below:
First Standard
Does the proposed amendment involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
As discussed in the licensee's application dated September 10,
2003, the VYNPS EPU analyses, which were performed at or above EPU
conditions, included a review and evaluation of the structures,
systems, and components (SSCs) that could be affected by the proposed
change. The licensee reviewed plant modifications and revised operating
parameters, including operator actions, to confirm acceptable
performance of plant SSCs under EPU conditions. On this basis, the
licensee concluded that there is no increase in the probability of
accidents previously evaluated.
Further, as also discussed in the licensee's application, while not
being submitted as a risk-informed licensing action, the proposed
amendment was evaluated by the licensee from a risk perspective. Using
the NRC guidelines established in Regulatory Guide (RG) 1.174, and the
calculated results from
[[Page 1775]]
the VYNPS Level 1 and 2 probabilistic safety analyses, the best
estimate for the core damage frequency (CDF) increase due to the
proposed EPU is 3.3 E-7 per year (an increase of 4.2 percent over the
pre-EPU CDF of 7.77 E-6 per year). The best estimate for the large
early release frequency (LERF) increase due to the proposed EPU is 1.1
E-7 per year (an increase of 4.9 percent over the pre-EPU LERF of 2.23
E-6 per year). The NRC staff concludes, based on review of the
licensee's risk evaluation and the acceptance guidelines in RG 1.174,
that the proposed amendment would not involve a significant increase in
the probability of an accident previously evaluated.
The NRC staff's evaluation of the proposed amendment included
review of the SSCs that could be affected by the proposed change. This
review included evaluation of plant modifications, revised operating
parameters, changes to operator actions and procedures, the EPU test
program, and changes to the plant TSs. Based on this review, the staff
concludes that there is reasonable assurance that the SSCs important to
safety will continue to meet their intended design basis functions
under EPU conditions. Therefore, the staff concludes that there is no
significant change in the ability of these SSCs to preclude or mitigate
the consequences of accidents.
The NRC staff's evaluation also reviewed the impact of the proposed
EPU on the radiological consequences of design-basis accidents for
VYNPS. The staff's review concluded that dose criteria in 10 CFR 50.67,
as well as the applicable acceptance criteria in Standard Review Plan
Section 15.0.1, would continue to be met at EPU conditions.
The NRC staff concludes, based on review of the SSCs that could be
affected by the proposed amendment and review of the radiological
consequences, that the proposed amendment would not involve a
significant increase in the consequences of an accident previously
evaluated.
Based on the above, the NRC staff concludes that the proposed
amendment would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
Second Standard
Does the proposed amendment create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
As stated above, the NRC staff's evaluation of the proposed
amendment included review of the SSCs that could be affected by the
proposed change. This review included evaluation of plant
modifications, revised operating parameters, changes to operator
actions and procedures, the EPU test program, and changes to the plant
TSs. Based on this review, the staff concludes that the proposed
amendment would not introduce any significantly new or different plant
equipment, would not significantly impact the manner in which the plant
is operated, and would not have any significant impact on the design
function or operation of the SCCs involved. The staff's review did not
identify any credible failure mechanisms, malfunctions, or accident
initiators not already considered in the VYNPS design and licensing
bases. Consequently, the staff concludes that the proposed change would
not introduce any failure mode not previously analyzed.
Based on the above, the NRC staff concludes that the proposed
change would not create the possibility of a new or different kind of
accident from any accident previously evaluated.
Third Standard
Does the proposed amendment involve a significant reduction in a
margin of safety?
Response: No.
As discussed in the licensee's application, continuing improvements
in analytical techniques based on several decades of boiling-water
reactor safety technology, plant performance feedback, operating
experience, and improved fuel and core designs, have resulted in a
significant increase in the design and operating margin between the
calculated safety analyses results and the current plant licensing
limits. The NRC staff's review found that the proposed EPU will reduce
some of the existing design and operational margins. However, safety
margins are considered to not be significantly reduced if: (1)
Applicable regulatory requirements, codes and standards or their
alternatives approved for use by the NRC, are met, and (2) if safety
analysis acceptance criteria in the licensing basis are met, or if
proposed revisions to the licensing basis provide sufficient margin to
account for analysis and data uncertainty.
Margin of safety is related to confidence in the ability of the
fission product barriers (i.e., fuel cladding, reactor coolant pressure
boundary (RCPB), and containment) to limit the level of radiation dose
to the public. The NRC staff evaluated the impact of the proposed EPU
on the fission product barriers as discussed below.
The NRC staff evaluated the impact of the proposed EPU to assure
that acceptable fuel damage limits are not exceeded. This included
consideration of the VYNPS fuel system design, nuclear system design,
thermal and hydraulic design, accident and transient analyses, and fuel
design limits. The evaluation included an assessment of the margin in
the associated safety analyses supporting the proposed EPU. The staff's
evaluation found that the licensee's analysis was acceptable based on
use of approved analytical methods and that the licensee had included
sufficient margin to account for analysis and data uncertainty. In
addition, the licensee will continue to perform cycle-specific analysis
to confirm that fuel design limits will not be exceeded during each
cycle. The staff's evaluation concluded that the applicable VYNPS
licensing basis requirements would continue to be met following
implementation of the proposed EPU (e.g., draft General Design Criteria
(GDC) 6, 7, and 8; and 10 CFR 50.46). Therefore, the NRC staff
concludes that fuel cladding integrity would be maintained within
acceptable limits under the proposed EPU conditions.
The NRC staff further evaluated the impact of the proposed EPU on
the RCPB. The evaluation included an assessment of overpressure
protection; structural integrity of the RCPB piping, components, and
supports; and structural integrity of the reactor vessel. With respect
to overpressure protection, the staff found that the licensee had used
an NRC-approved evaluation method, had used the most limiting
pressurization event, and had determined that the peak calculated
pressure would remain below the American Society of Mechanical
Engineers Boiler and Pressure Vessel Code (ASME Code) allowable peak
pressure. With respect to structural integrity of the RCPB piping,
components, and supports, the staff found that the licensee had
performed its evaluation using the process and methodology defined in
NRC-approved topical reports. The staff's evaluation concluded that
RCPB structural integrity would be maintained at EPU conditions. With
respect to structural integrity of the reactor vessel, the staff found
that the licensee had implemented an acceptable reactor vessel
materials surveillance program in a previously approved amendment that
was based on neutron fluence values acceptable for VYNPS at EPU
conditions. In addition, the staff found that the existing pressure-
temperature limit curves contained in the TSs would remain
[[Page 1776]]
bounding for EPU conditions. The staff also found that the methodology
used by the licensee to evaluate the loads on the reactor vessel was
consistent with an NRC-approved methodology and that the maximum
stresses and fatigue usage factors for EPU conditions would be within
ASME Code allowable limits. The staff's evaluation regarding the RCPB
concluded that the applicable VYNPS licensing basis requirements would
continue to be met following implementation of the proposed EPU (e.g.,
draft GDC 9, 33, 34, and 35; 10 CFR 50.60; and 10 CFR part 50,
Appendices G and H). Therefore, the NRC staff concludes that RCPB
structural integrity would be maintained under the proposed EPU
conditions.
Finally, the NRC staff evaluated the impact of the proposed EPU on
the containment. The staff found that the licensee's analysis used
acceptable calculational methods and conservative assumptions and that
the containment pressure and temperature under EPU conditions would
remain below existing design limits. The staff also evaluated the
licensee's proposed change to the licensing basis to credit containment
accident pressure to meet the net positive suction head (NPSH)
requirements for the emergency core cooling system pumps. The staff
found that the licensee's analysis was performed using conservative
assumptions and that the credited pressure remains below the
containment accident pressure that would be available under EPU
conditions. The staff's evaluation regarding the containment concluded
that the applicable VYNPS licensing basis requirements would continue
to be met following implementation of the proposed EPU (e.g., draft GDC
10, 41, 49, and 52; and 10 CFR part 50, Appendix K). Therefore, the NRC
staff concludes that containment structural integrity would be
maintained under the proposed EPU conditions.
In summary, the NRC staff has concluded that the structural
integrity of the fission product barriers (i.e., fuel cladding, RCPB
and containment) would be maintained under EPU conditions. As such, the
proposed amendment would not degrade confidence in the ability of the
barriers to limit the level of radiation dose to the public.
Based on the above, the NRC staff concludes that the proposed
change would not involve a significant reduction in a margin of safety.
Conclusion
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 a final
determination.
The Commission previously published a ``Notice of Consideration of
Issuance of Amendment to Facility Operating License and Opportunity for
a Hearing'' for the proposed VYNPS EPU amendment in the Federal
Register on July 1, 2004 (69 FR 39976). This Notice provided 60 days
for the public to request a hearing. On August 30, 2004, the Vermont
Department of Public Service and the New England Coalition filed
requests for hearing in connection with the proposed amendment. By
Order dated November 22, 2004, the Atomic Safety and Licensing Board
(ASLB) granted those hearing requests and by Order dated December 16,
2004, the ASLB issued its decision to conduct a hearing using the
procedures in 10 CFR part 2, subpart L, ``Informal Hearing Procedures
for NRC Adjudications.'' No additional opportunity for hearing is
provided in connection with this notice.
In accordance with the Commission's regulations in 10 CFR 50.91, if
a final determination is made that the proposed amendment involves no
significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding submission
of adverse comments or a request for hearing. In that event, any
required hearing would be completed after issuance of the amendment;
however, if a final determination is made that the proposed amendment
involves a significant hazards consideration, the amendment would not
be issued prior to completion of the hearing.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administrative 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 6D59, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays.
For further details with respect to the proposed action, see the
licensee's application dated September 10, 2003, as supplemented on
October 1, and October 28 (2 letters), 2003, January 31 (2 letters),
March 4, May 19, July 2, July 27, July 30, August 12, August 25,
September 14, September 15, September 23, September 30 (2 letters),
October 5, October 7 (2 letters), December 8, and December 9, 2004, and
February 24, March 10, March 24, March 31, April 5, April 22, June 2,
August 1, August 4, September 10, September 14, September 18, September
28, October 17, October 21, 2005 (2 letters), October 26, October 29,
November 2, November 22, and December 2, 2005. Documents may be
examined, and/or copied for a fee, at the NRC's Public Document Room
(PDR), located at One White Flint North, Public File Area O1 F21, 11555
Rockville Pike (first floor), Rockville, Maryland. Publicly available
records will be accessible electronically from the ADAMS Public
Electronic Reading Room on the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. Persons who do not have access to ADAMS or who
encounter problems in accessing the documents located in ADAMS should
contact the NRC PDR Reference staff at 1-800-397-4209, or 301-415-4737,
or send an e-mail to [email protected].
Dated at Rockville, Maryland, this 5th day of January 2006.
For the Nuclear Regulatory Commission.
Richard B. Ennis,
Senior Project Manager, Plant Licensing Branch I-2, Division of
Operating Reactor Licensing, Office of Nuclear Reactor Regulation.
[FR Doc. E6-159 Filed 1-10-06; 8:45 am]
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