[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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