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, 1774-1776 [E6-159]
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Federal Register / Vol. 71, No. 7 / Wednesday, January 11, 2006 / Notices
comparable CES all employees figure,
temporary help agency workers, leased
workers, independent contractors, and
other workers not classified elsewhere.
The BLS plans to re-contact 100 of the
16,000 respondents to verify the quality
of the responses received.
Reporting for the CES survey is
voluntary under federal law, but is
mandatory under state law in five
States. The supplemental survey will
not be using the State mandatory
reporting authority.
The BLS may conduct additional
supplemental surveys in the future,
depending on the availability of
resources and the significance of the
topic. The BLS is requesting approval
for collection through December 31,
2006.
• Enhance the quality, utility, and
clarity of the information to be
collected; and
• Minimize the burden of the
collection of information on those who
are to respond, including through the
use of appropriate automated,
electronic, mechanical, or other
technological collection techniques or
other forms of information technology,
e.g., permitting electronic submissions
of responses.
Type of Review: New Collection.
Agency: Bureau of Labor Statistics.
Title: CES Supplemental Form on
Temporary Help, Leased, and Other
Contracted Work.
OMB Number: 1220–NEW.
Affected Public: Businesses or other
for-profit; Small businesses or
organizations.
II. Current Action
Office of Management and Budget
Clearance is being sought for the CES
Supplemental Form on Temporary
Help, Leased, and Other Contracted
Work.
III. Desired Focus of Comments
The BLS is particularly interested in
comments that:
• Evaluate whether the proposed
collection of information is necessary
for the proper performance of the
functions of the agency, including
whether the information will have
practical utility;
• Evaluate the accuracy of the
agency’s estimate of the burden of the
proposed collection of information,
including the validity of the
methodology and assumptions used;
Number of
respondents
Form
Minutes per
report
Frequency of
response
Annual
responses
Annual burden
hours
Supplemental Form on Temporary Help, Leased, and
Other Contracted Work ....................................................
Response Analysis interviews .............................................
16,000
100
20
120
1
1
16,000
100
5,333
200
Total ..............................................................................
16,100
........................
........................
16,100
5,533
Total Burden Cost (capital/startup):
$0.
Total Burden Cost (operating/
maintenance): $0.
Comments submitted in response to
this notice will be summarized and/or
included in the request for Office of
Management and Budget approval of the
information collection request; they also
will become a matter of public record.
Signed at Washington, DC, this 3rd day of
January 2006.
Kimberley Hill,
Acting Chief, Division of Management
Systems, Bureau of Labor Statistics.
[FR Doc. E6–149 Filed 1–10–06; 8:45 am]
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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
VerDate Aug<31>2005
14:20 Jan 10, 2006
Jkt 208001
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), § 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
PO 00000
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Fmt 4703
Sfmt 4703
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
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Federal Register / Vol. 71, No. 7 / Wednesday, January 11, 2006 / Notices
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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
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Jkt 208001
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.
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Fmt 4703
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1775
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 cyclespecific 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
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Federal Register / Vol. 71, No. 7 / Wednesday, January 11, 2006 / Notices
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)
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14:20 Jan 10, 2006
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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
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Fmt 4703
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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, https://
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
pdr@nrc.gov.
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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Radiation Source Protection and
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Nuclear Regulatory
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AGENCY:
ACTION:
Request for public comment.
SUMMARY: The Nuclear Regulatory
Commission (NRC) has established an
interagency task force to evaluate and
make recommendations on the
protection and security of radiation
sources. The Radiation Source
Protection and Security Task Force
(Task Force) is required by the Energy
Policy Act of 2005. As part of the Task
Force’s considerations, it is seeking
public input on the major issues before
the Task Force. To aid in that process,
the NRC is requesting comments on the
issues discussed in this notice.
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]]>Agencies
[Federal Register Volume 71, Number 7 (Wednesday, January 11, 2006)]
[Notices]
[Pages 1774-1776]
From the Federal Register Online via the Government Printing 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, https://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 pdr@nrc.gov.
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]
BILLING CODE 7590-01-P
