Arizona Public Service Company, et al.; Palo Verde Nuclear Generating Station, Units 1, 2, and 3; Exemption, 9623-9625 [2010-4388]
Download as PDF
Federal Register / Vol. 75, No. 41 / Wednesday, March 3, 2010 / Notices
Nuclear Energy Institute). The licensee’s
request for an exemption is therefore
consistent with the approach set forth
by the Commission as discussed in the
June 4, 2009, letter.
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VEGP Schedule Exemption Request
The licensee provided detailed
information in its letter dated November
6, 2009, as supplemented November 20,
2009, requesting an exemption. It
describes a comprehensive plan to
install equipment related to a certain
requirement in the new Part 73 rule and
provides a timeline for achieving full
compliance with the new regulation.
The submittals contain proprietary
information regarding the site security
plan, details of the specific requirement
of the regulation for which the site
cannot be in compliance by the March
31, 2010, deadline and why, the
required changes to the site’s security
configuration, and a timeline with
critical path activities that will bring the
licensee into full compliance by
September 27, 2010. The timeline
provides dates indicating (1) when
various phases of the project begin and
end (i.e., design, field construction), (2)
outages scheduled for each unit, and (3)
when critical equipment will be
ordered, installed, tested and become
operational.
Notwithstanding the schedular
exemption for this limited requirement,
the licensee will continue to be in
compliance with all other applicable
physical security requirements as
described in 10 CFR 73.55 and reflected
in its current NRC approved physical
security program. By September 27,
2010, VEGP will be in full compliance
with all the regulatory requirements of
10 CFR 73.55, as issued on March 27,
2009.
4.0 Conclusion for Part 73 Schedule
Exemption Request
The NRC staff has reviewed the
licensee’s submittals and concludes that
the licensee has provided adequate
justification for its request for an
extension of the compliance date to
September 27, 2010, with regard to a
specific requirement of 10 CFR 73.55.
Accordingly, the Commission has
determined that pursuant to10 CFR
73.5, ‘‘Specific exemptions,’’ an
exemption from the March 31, 2010,
compliance date is authorized by law
and will not endanger life or property or
the common defense and security, and
is otherwise in the public interest.
Therefore, the Commission hereby
grants the requested exemption.
The NRC staff has determined that the
long-term benefits that will be realized
when the VEGP equipment installation
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is complete justifies extending the full
compliance date with regard to the
specific requirement of 10 CFR 73.55.
The security measure, that VEGP needs
additional time to implement, is a new
requirement imposed by the March 27,
2009, amendments to 10 CFR 73.55, and
is in addition to those required by the
security orders issued in response to the
events of September 11, 2001.
Therefore, it is concluded that the
licensee’s actions are in the best interest
of protecting the public health and
safety through the security changes that
will result from granting this exemption.
As per the licensee’s request and the
NRC’s regulatory authority to grant an
exemption from the March 31, 2010,
implementation deadline for the
requirement specified in the SNC letter
dated November 6, 2009, as
supplemented November 20, 2009, the
licensee is required to be in full
compliance by September 27, 2010. In
achieving compliance, the licensee is
reminded that it is responsible for
determining the appropriate licensing
mechanism (i.e., 10 CFR 50.54(p) or 10
CFR 50.90) for incorporation of all
necessary changes to its security plans.
Pursuant to 10 CFR 51.32, ‘‘Finding of
no significant impact,’’ the Commission
has previously determined that the
granting of this exemption will not have
a significant effect on the quality of the
human environment 75 FR 3943; dated
January 25, 2010.
This exemption is effective upon
issuance.
Dated at Rockville, Maryland, this 24 day
of February 2010.
For the Nuclear Regulatory Commission.
Allen G. Howe,
Acting Director, Division of Operating Reactor
Licensing, Office of Nuclear Reactor
Regulation.
[FR Doc. 2010–4381 Filed 3–2–10; 8:45 am]
BILLING CODE 7590–01–P
NUCLEAR REGULATORY
COMMISSION
[Docket Nos. STN 50–528, STN 50–529, and
STN 50–530; NRC–2010–0058]
Arizona Public Service Company, et
al.; Palo Verde Nuclear Generating
Station, Units 1, 2, and 3; Exemption
1.0 Background
The Arizona Public Service Company
(APS, the facility licensee) is the holder
of Facility Operating License Nos. NPF–
41, NPF–51, and NPF–74, which
authorize operation of the Palo Verde
Nuclear Generating Station (PVNGS, the
facility), Units 1, 2, and 3, respectively.
The licenses provide, among other
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9623
things, that the facility is subject to all
rules, regulations, and orders of the
Nuclear Regulatory Commission (NRC,
or the Commission) now or hereafter in
effect.
The facility consists of three
pressurized-water reactors located in
Maricopa County, Arizona.
2.0 Request/Action
Title 10 of the Code of Federal
Regulations (10 CFR) Part 50, Appendix
G, ‘‘Fracture Toughness Requirements,’’
which is invoked by 10 CFR 50.60,
requires that pressure-temperature (P–T)
limits be established for the reactor
coolant pressure boundary during
normal operating and hydrostatic or
leak rate testing conditions. Specifically,
10 CFR Part 50, Appendix G states that
‘‘[t]he appropriate requirements on both
the pressure-temperature limits and the
minimum permissible temperature must
be met for all conditions,’’ and ‘‘[t]he
pressure-temperature limits identified
as ‘ASME [American Society for
Mechanical Engineers] Appendix G
limits’ in Table 3 require that the limits
must be at least as conservative as limits
obtained by following the methods of
analysis and the margins of safety of
Appendix G of Section XI of the ASME
Code [Boiler and Pressure Vessel
Code].’’ The regulations in 10 CFR Part
50, Appendix G, also specify the
applicable editions and addenda of the
ASME Code, Section XI, which are
incorporated by reference in 10 CFR
50.55a. In the most recent version of 10
CFR (2009 Edition), the 1977 Edition
through the 2004 Edition of the ASME
Code, Section XI are incorporated by
reference in 10 CFR 50.55a. Finally, 10
CFR 50.60(b) states that, ‘‘[p]roposed
alternatives to the described
requirements in Append[ix] G * * * of
this part or portions thereof may be used
when an exemption is granted by the
Commission under [10 CFR] 50.12.’’
By letter dated February 19, 2009, as
supplemented by letter dated December
22, 2009 (Agencywide Documents
Access and Management System
(ADAMS) Accession Nos. ML090641014
and ML100040069, respectively), the
licensee submitted a request for
exemption from 10 CFR Part 50,
Appendix G regarding the pressuretemperature (P–T) limits calculation,
and a license amendment request to
revise Technical Specification (TS) 3.4,
‘‘Reactor Coolant System (RCS),’’ to
relocate the P–T limits and the low
temperature overpressure protection
(LTOP) system enable temperatures
from the TS to a licensee-controlled
document; the Pressure and
Temperature Limits Report (PTLR). In
the license amendment request, the
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03MRN1
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9624
Federal Register / Vol. 75, No. 41 / Wednesday, March 3, 2010 / Notices
licensee identified Combustion
Engineering (CE) Owners Group Topical
Report CE NPSD–683–A, Revision 6,
‘‘Development of a RCS Pressure and
Temperature Limits Report (PTLR) for
the Removal of P–T Limits and LTOP
Requirements from the Technical
Specifications’’ (ADAMS Accession No.
ML011350387), as the PTLR
methodology that would be cited in the
administrative controls section of the
PVNGS, Units 1, 2, and 3 Technical
Specifications governing PTLR content.
The NRC staff evaluated the specific
PTLR methodology in CE NPSD–683,
Revision 6. This evaluation was
documented in the NRC safety
evaluation (SE) of March 16, 2001
(ADAMS Accession No. ML010780017),
which specified additional licensee
actions that are necessary to support a
licensee’s adoption of CE NPSD–683,
Revision 6. The final approved version
of this report was reissued as CE NPSD–
683–A, Revision 6, which included the
NRC SE and the required additional
action items as an attachment to the
report. One of the additional specified
actions stated that if a licensee proposed
to use the methodology in CE NPSD–
683–A, Revision 6, for the calculation of
flaw stress intensity factors due to
membrane stress from pressure loading
(KIM), an exemption was required, since
the methodology for the calculation of
KIM values in CE NPSD–683–A,
Revision 6, could not be shown to be
conservative with respect to the
methodology for the determination of
KIM provided in editions and addenda
of the ASME Code, Section XI,
Appendix G through the 2004 Edition.
Therefore, in addition to the license
amendment request, the licensee’s
February 19, 2009, submittal also
contains an exemption request,
consistent with the requirements of 10
CFR 50.12 and 50.60, to apply the KIM
calculational methodology of CE NPSD–
683–A, Revision 6, as part of the
PVNGS, Units 1, 2, and 3 PTLR
methodology.
During the NRC staff’s review of CE
NPSD–683, Revision 6, the NRC staff
evaluated the KIM calculational
methodology of that report versus the
methodologies for the calculation of KIM
given in the ASME Code, Section XI,
Appendix G. In the NRC’s March 16,
2001, SE, the staff noted, ‘‘[t]he CE NSSS
[nuclear steam supply system]
methodology does not invoke the
methods in the 1995 edition of
Appendix G to the Code for calculating
KIM factors, and instead applies FEM
[finite element modeling] methods for
estimating the KIM factors for the RPV
[reactor pressure vessel] shell * * * the
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16:08 Mar 02, 2010
Jkt 220001
staff has determined that the KIM
calculation methods apply FEM
modeling that is similar to that used for
the determination of the KIT factors [as
codified in the ASME Code, Section XI,
Appendix G]. The staff has also
determined that there is only a slight
non-conservative difference between the
P–T limits generated from the 1989
edition of Appendix G to the Code and
those generated from CE NSSS
methodology as documented in CE/ABB
Evaluation 063–PENG–ER–096,
Revision 00, ‘‘Technical Methodology
Paper Comparing ABB/CE PT Curve to
ASME Section III, Appendix G,’’ dated
January 22, 1998 (ADAMS Accession
No. ML100500514, non-proprietary
version). The staff considers that this
difference is reasonable and that it will
be consistent with the expected
improvements in P–T generation
methods that have been incorporated
into the 1995 edition of Appendix G to
the Code.’’ This conclusion regarding
the comparison between the CE NSSS
methodology and the 1995 Edition of
the ASME Code, Section XI, Appendix
G methodology also applies to the 2004
Edition of the ASME Code, Section XI,
Appendix G methodology because the
evolution of the ASME Code Section XI,
Appendix G methodology does not
affect the KIM calculation significantly.
In summary, the staff concluded in its
March 16, 2001, SE that the calculation
of KIM using the CE NPSD–683, Revision
6 methodology would lead to the
development of P–T limit curves which
may be slightly non-conservative with
respect to those which would be
calculated using the ASME Code,
Section XI, Appendix G methods, and
that such a difference was to be
expected with the development of more
refined calculational techniques.
Furthermore, the staff concluded in its
March 16, 2001, SE that P–T limit
curves that would be developed using
the methodology of CE NPSD–683,
Revision 6 would be adequate for
protecting the RPV from brittle fracture
under all normal operating and
hydrostatic/leak test conditions.
3.0 Discussion
Pursuant to 10 CFR 50.12, the
Commission may, upon application by
any interested person or upon its own
initiative, grant exemptions from the
requirements of 10 CFR Part 50 when (1)
the exemptions are authorized by law,
will not present an undue risk to public
health or safety, and are consistent with
the common defense and security; and
(2) when special circumstances are
present.
This exemption results in changes to
the plant by allowing the use of an
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Fmt 4703
Sfmt 4703
alternative methodology for calculating
flaw stress intensity factors in the RPV
due to membrane stress from pressure
loadings in lieu of meeting the
requirements in 10 CFR 50.60 and 10
CFR Part 50, Appendix G. As stated
above, 10 CFR 50.12 allows NRC to
grant exemptions from the requirements
of 10 CFR Part 50. In addition, the
granting of the exemption will not result
in violation of the Atomic Energy Act of
1954, as amended, or the Commission’s
regulations. Therefore, the exemption is
authorized by law.
The underlying purpose of 10 CFR
50.60 and 10 CFR part 50, appendix G
is to ensure that appropriate P–T limits
and the minimum permissible
temperature are established for the RPV
under normal operating and hydrostatic
or leak rate test conditions. The
licensee’s alternative methodology for
establishing the P–T limits and the
LTOP setpoints is described in CE
NPSD–683–A, Revision 6, which has
been approved by the NRC staff. Based
on the above, no new accident
precursors are created by using the
alternative methodology. Thus, the
probability of postulated accidents is
not increased. Also, based on the above,
the consequences of postulated
accidents are not increased. In addition,
the licensee used an NRC-approved
methodology for establishing P–T limits
and minimum permissible temperatures
for the reactor vessel. Therefore, there is
no undue risk to the public health and
safety.
The exemption results in changes to
the plant by allowing an alternative
methodology for calculating flaw stress
intensity factors in the reactor vessel.
This change to the calculation of stress
intensity factors in the reactor vessel
material has no negative implications
for security issues. Therefore, the
common defense and security is not
impacted by this exemption.
Special circumstances, pursuant to 10
CFR 50.12(a)(2)(ii), are present in that
continued operation of PVNGS, Units 1,
2, and 3 with P–T limit curves
developed in accordance with the
ASME Code, Section XI, Appendix G is
not necessary to achieve the underlying
purpose of 10 CFR part 50, appendix G.
Application of the KIM calculational
methodology of CE NPSD–683–A,
Revision 6 in lieu of the calculational
methodology specified in the ASME
Code, Section XI, Appendix G provides
an acceptable alternative evaluation
procedure, which will continue to meet
the underlying purpose of 10 CFR part
50, appendix G. The underlying purpose
of the regulations in 10 CFR part 50,
appendix G is to provide an acceptable
margin of safety against brittle failure of
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Federal Register / Vol. 75, No. 41 / Wednesday, March 3, 2010 / Notices
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the reactor coolant system during any
condition of normal operation to which
the pressure boundary may be subjected
over its service lifetime.
Based on the staff’s March 16, 2001,
SE regarding CE NPSD–683, Revision 6
and the licensee’s rationale to support
the exemption request, the staff agrees
with the licensee’s determination that
an exemption is required to approve the
use of the KIM calculational
methodology of CE NPSD–683–A,
Revision 6. The staff concludes that the
application of the KIM calculational
methodology of CE NPSD–683–A,
Revision 6, for PVNGS, Units 1, 2, and
3 provides sufficient margin in the
development of RPV P–T limit curves
such that the underlying purpose of the
regulations (10 CFR part 50, appendix
G) continues to be met. Therefore, the
NRC staff concludes that the exemption
requested by the licensee is justified
based on the special circumstances of 10
CFR 50.12(a)(2)(ii), ‘‘[a]pplication of the
regulation in the particular
circumstances would not serve the
underlying purpose of the rule or is not
necessary to achieve the underlying
purpose of the rule.’’
Based upon a consideration of the
conservatism that is incorporated into
the methodologies of 10 CFR part 50,
appendix G and ASME Code, Section
XI, Appendix G, the staff concludes that
application of the KIM calculational
methodology of CE NPSD–683–A,
Revision 6, as described, would provide
an adequate margin of safety against
brittle failure of the RPV. Therefore, the
staff concludes that the exemption is
appropriate under the special
circumstances of 10 CFR 50.12(a)(2)(ii),
and that the application of the KIM
calculational methodology of CE NPSD–
683–A, Revision 6, is acceptable for use
in the PVNGS, Units 1, 2, and 3 PTLR
methodology.
4.0 Conclusion
Accordingly, the Commission has
determined that, pursuant to 10 CFR
50.12(a), the exemption is authorized by
law, will not present an undue risk to
the public health and safety, and is
consistent with the common defense
and security. Also, special
circumstances are present. Therefore,
the Commission hereby grants APS an
exemption from the requirements of 10
CFR part 50, appendix G to allow
application of the KIM calculational
methodology of CE NPSD–683–A,
Revision 6 in establishing the PTLR
methodology for PVNGS, Units 1, 2, and
3.
Pursuant to 10 CFR 51.32, the
Commission has determined that the
granting of this exemption will not have
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16:08 Mar 02, 2010
Jkt 220001
a significant effect on the quality of the
human environment (75 FR 8149; dated
February 23, 2010).
This exemption is effective upon
issuance.
Dated at Rockville, Maryland, this 24th day
of February 2010.
For the Nuclear Regulatory Commission.
Allen G. Howe,
Acting Director, Division of Operating Reactor
Licensing, Office of Nuclear Reactor
Regulation.
[FR Doc. 2010–4388 Filed 3–2–10; 8:45 am]
BILLING CODE 7590–01–P
NUCLEAR REGULATORY
COMMISSION
[Docket Nos. 50–282 and 50–306; NRC–
2010–0046]
Northern States Power Company—
Minnesota Prairie Island Nuclear
Generating Plant, Units 1 and 2;
Exemption
1.0
Background
Northern States Power Company, a
Minnesota corporation (NSPM, the
licensee) is the holder of Facility
Operating License Nos. DPR–42 and
DPR–60, which authorize operation of
the Prairie Island Nuclear Generating
Plant, Units 1 and 2 (PINGP). The
licenses provide, among other things,
that the facility is subject to all rules,
regulations, and orders of the U.S.
Nuclear Regulatory Commission (NRC,
the Commission) now or hereafter in
effect.
The facility consists of two
pressurized-water reactors located in
Goodhue County, Minnesota.
2.0
Request/Action
Title 10 of the Code of Federal
Regulations (10 CFR) Part 73, ‘‘Physical
protection of plants and materials,’’
Section 73.55, ‘‘Requirements for
physical protection of licensed activities
in nuclear power reactors against
radiological sabotage,’’ published March
27, 2009, effective May 26, 2009, with
a full implementation date of March 31,
2010, requires licensees to protect, with
high assurance, against radiological
sabotage by designing and
implementing comprehensive site
security programs. The amendments to
10 CFR 73.55 published on March 27,
2009, establish and update generically
applicable security requirements similar
to those previously imposed by
Commission orders issued after the
terrorist attacks of September 11, 2001
and implemented by licensees. In
addition, the amendments to 10 CFR
73.55 include additional requirements
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Sfmt 4703
9625
to further enhance site security based
upon insights gained from
implementation of the post September
11, 2001 security orders. It is from five
of these new requirements that PINGP
now seeks an exemption from the March
31, 2010 implementation date. All other
physical security requirements
established by this recent rulemaking
have already been or will be
implemented by the licensee by March
31, 2010.
By letter dated November 5, 2009, as
supplemented by letters dated
November 30 and December 17, 2009,
the licensee requested an exemption in
accordance with 10 CFR 73.5, ‘‘Specific
exemptions.’’ The licensee’s November 5
and December 17, 2009, letters, and
certain portions of its November 30,
2009 (Agencywide Documents Access
and Management System (ADAMS)
Accession No. ML100050096), letter,
contain security-related information
and, accordingly, are not available to the
public. The licensee has requested an
exemption from the March 31, 2010,
compliance date identified in 10 CFR
73.55(a)(1), stating that specific parts of
the new requirements will require more
time to implement before all
requirements can be met. Specifically,
the request is to extend the compliance
date for five specific requirements from
the current March 31, 2010, deadline to
June 30, 2011. Being granted this
exemption for the five requirements
would allow the licensee to complete
the modifications designed to provide
significant upgrades to the security
system to meet the noted regulatory
requirements.
3.0 Discussion of Part 73 Schedule
Exemptions From the March 31, 2010,
Full Implementation Date
Pursuant to 10 CFR 73.55(a)(1), ‘‘By
March 31, 2010, each nuclear power
reactor licensee, licensed under 10 CFR
part 50, shall implement the
requirements of this section through its
Commission-approved Physical Security
Plan, Training and Qualification Plan,
Safeguards Contingency Plan, and Cyber
Security Plan, referred to collectively
hereafter as ‘security plans.’ ’’ Pursuant
to 10 CFR 73.5, the Commission may,
upon application by any interested
person or upon its own initiative, grant
exemptions from the requirements of 10
CFR part 73 when the exemptions are
authorized by law, and will not
endanger life or property or the common
defense and security, and are otherwise
in the public interest.
NRC approval of this exemption
would, as noted above, extend the
required compliance date for the
requirements specified in the licensee’s
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Agencies
[Federal Register Volume 75, Number 41 (Wednesday, March 3, 2010)]
[Notices]
[Pages 9623-9625]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2010-4388]
-----------------------------------------------------------------------
NUCLEAR REGULATORY COMMISSION
[Docket Nos. STN 50-528, STN 50-529, and STN 50-530; NRC-2010-0058]
Arizona Public Service Company, et al.; Palo Verde Nuclear
Generating Station, Units 1, 2, and 3; Exemption
1.0 Background
The Arizona Public Service Company (APS, the facility licensee) is
the holder of Facility Operating License Nos. NPF-41, NPF-51, and NPF-
74, which authorize operation of the Palo Verde Nuclear Generating
Station (PVNGS, the facility), Units 1, 2, and 3, respectively. The
licenses provide, among other things, that the facility is subject to
all rules, regulations, and orders of the Nuclear Regulatory Commission
(NRC, or the Commission) now or hereafter in effect.
The facility consists of three pressurized-water reactors located
in Maricopa County, Arizona.
2.0 Request/Action
Title 10 of the Code of Federal Regulations (10 CFR) Part 50,
Appendix G, ``Fracture Toughness Requirements,'' which is invoked by 10
CFR 50.60, requires that pressure-temperature (P-T) limits be
established for the reactor coolant pressure boundary during normal
operating and hydrostatic or leak rate testing conditions.
Specifically, 10 CFR Part 50, Appendix G states that ``[t]he
appropriate requirements on both the pressure-temperature limits and
the minimum permissible temperature must be met for all conditions,''
and ``[t]he pressure-temperature limits identified as `ASME [American
Society for Mechanical Engineers] Appendix G limits' in Table 3 require
that the limits must be at least as conservative as limits obtained by
following the methods of analysis and the margins of safety of Appendix
G of Section XI of the ASME Code [Boiler and Pressure Vessel Code].''
The regulations in 10 CFR Part 50, Appendix G, also specify the
applicable editions and addenda of the ASME Code, Section XI, which are
incorporated by reference in 10 CFR 50.55a. In the most recent version
of 10 CFR (2009 Edition), the 1977 Edition through the 2004 Edition of
the ASME Code, Section XI are incorporated by reference in 10 CFR
50.55a. Finally, 10 CFR 50.60(b) states that, ``[p]roposed alternatives
to the described requirements in Append[ix] G * * * of this part or
portions thereof may be used when an exemption is granted by the
Commission under [10 CFR] 50.12.''
By letter dated February 19, 2009, as supplemented by letter dated
December 22, 2009 (Agencywide Documents Access and Management System
(ADAMS) Accession Nos. ML090641014 and ML100040069, respectively), the
licensee submitted a request for exemption from 10 CFR Part 50,
Appendix G regarding the pressure-temperature (P-T) limits calculation,
and a license amendment request to revise Technical Specification (TS)
3.4, ``Reactor Coolant System (RCS),'' to relocate the P-T limits and
the low temperature overpressure protection (LTOP) system enable
temperatures from the TS to a licensee-controlled document; the
Pressure and Temperature Limits Report (PTLR). In the license amendment
request, the
[[Page 9624]]
licensee identified Combustion Engineering (CE) Owners Group Topical
Report CE NPSD-683-A, Revision 6, ``Development of a RCS Pressure and
Temperature Limits Report (PTLR) for the Removal of P-T Limits and LTOP
Requirements from the Technical Specifications'' (ADAMS Accession No.
ML011350387), as the PTLR methodology that would be cited in the
administrative controls section of the PVNGS, Units 1, 2, and 3
Technical Specifications governing PTLR content. The NRC staff
evaluated the specific PTLR methodology in CE NPSD-683, Revision 6.
This evaluation was documented in the NRC safety evaluation (SE) of
March 16, 2001 (ADAMS Accession No. ML010780017), which specified
additional licensee actions that are necessary to support a licensee's
adoption of CE NPSD-683, Revision 6. The final approved version of this
report was reissued as CE NPSD-683-A, Revision 6, which included the
NRC SE and the required additional action items as an attachment to the
report. One of the additional specified actions stated that if a
licensee proposed to use the methodology in CE NPSD-683-A, Revision 6,
for the calculation of flaw stress intensity factors due to membrane
stress from pressure loading (KIM), an exemption was
required, since the methodology for the calculation of KIM
values in CE NPSD-683-A, Revision 6, could not be shown to be
conservative with respect to the methodology for the determination of
KIM provided in editions and addenda of the ASME Code,
Section XI, Appendix G through the 2004 Edition. Therefore, in addition
to the license amendment request, the licensee's February 19, 2009,
submittal also contains an exemption request, consistent with the
requirements of 10 CFR 50.12 and 50.60, to apply the KIM
calculational methodology of CE NPSD-683-A, Revision 6, as part of the
PVNGS, Units 1, 2, and 3 PTLR methodology.
During the NRC staff's review of CE NPSD-683, Revision 6, the NRC
staff evaluated the KIM calculational methodology of that
report versus the methodologies for the calculation of KIM
given in the ASME Code, Section XI, Appendix G. In the NRC's March 16,
2001, SE, the staff noted, ``[t]he CE NSSS [nuclear steam supply
system] methodology does not invoke the methods in the 1995 edition of
Appendix G to the Code for calculating KIM factors, and
instead applies FEM [finite element modeling] methods for estimating
the KIM factors for the RPV [reactor pressure vessel] shell
* * * the staff has determined that the KIM calculation
methods apply FEM modeling that is similar to that used for the
determination of the KIT factors [as codified in the ASME
Code, Section XI, Appendix G]. The staff has also determined that there
is only a slight non-conservative difference between the P-T limits
generated from the 1989 edition of Appendix G to the Code and those
generated from CE NSSS methodology as documented in CE/ABB Evaluation
063-PENG-ER-096, Revision 00, ``Technical Methodology Paper Comparing
ABB/CE PT Curve to ASME Section III, Appendix G,'' dated January 22,
1998 (ADAMS Accession No. ML100500514, non-proprietary version). The
staff considers that this difference is reasonable and that it will be
consistent with the expected improvements in P-T generation methods
that have been incorporated into the 1995 edition of Appendix G to the
Code.'' This conclusion regarding the comparison between the CE NSSS
methodology and the 1995 Edition of the ASME Code, Section XI, Appendix
G methodology also applies to the 2004 Edition of the ASME Code,
Section XI, Appendix G methodology because the evolution of the ASME
Code Section XI, Appendix G methodology does not affect the
KIM calculation significantly.
In summary, the staff concluded in its March 16, 2001, SE that the
calculation of KIM using the CE NPSD-683, Revision 6
methodology would lead to the development of P-T limit curves which may
be slightly non-conservative with respect to those which would be
calculated using the ASME Code, Section XI, Appendix G methods, and
that such a difference was to be expected with the development of more
refined calculational techniques. Furthermore, the staff concluded in
its March 16, 2001, SE that P-T limit curves that would be developed
using the methodology of CE NPSD-683, Revision 6 would be adequate for
protecting the RPV from brittle fracture under all normal operating and
hydrostatic/leak test conditions.
3.0 Discussion
Pursuant to 10 CFR 50.12, the Commission may, upon application by
any interested person or upon its own initiative, grant exemptions from
the requirements of 10 CFR Part 50 when (1) the exemptions are
authorized by law, will not present an undue risk to public health or
safety, and are consistent with the common defense and security; and
(2) when special circumstances are present.
This exemption results in changes to the plant by allowing the use
of an alternative methodology for calculating flaw stress intensity
factors in the RPV due to membrane stress from pressure loadings in
lieu of meeting the requirements in 10 CFR 50.60 and 10 CFR Part 50,
Appendix G. As stated above, 10 CFR 50.12 allows NRC to grant
exemptions from the requirements of 10 CFR Part 50. In addition, the
granting of the exemption will not result in violation of the Atomic
Energy Act of 1954, as amended, or the Commission's regulations.
Therefore, the exemption is authorized by law.
The underlying purpose of 10 CFR 50.60 and 10 CFR part 50, appendix
G is to ensure that appropriate P-T limits and the minimum permissible
temperature are established for the RPV under normal operating and
hydrostatic or leak rate test conditions. The licensee's alternative
methodology for establishing the P-T limits and the LTOP setpoints is
described in CE NPSD-683-A, Revision 6, which has been approved by the
NRC staff. Based on the above, no new accident precursors are created
by using the alternative methodology. Thus, the probability of
postulated accidents is not increased. Also, based on the above, the
consequences of postulated accidents are not increased. In addition,
the licensee used an NRC-approved methodology for establishing P-T
limits and minimum permissible temperatures for the reactor vessel.
Therefore, there is no undue risk to the public health and safety.
The exemption results in changes to the plant by allowing an
alternative methodology for calculating flaw stress intensity factors
in the reactor vessel. This change to the calculation of stress
intensity factors in the reactor vessel material has no negative
implications for security issues. Therefore, the common defense and
security is not impacted by this exemption.
Special circumstances, pursuant to 10 CFR 50.12(a)(2)(ii), are
present in that continued operation of PVNGS, Units 1, 2, and 3 with P-
T limit curves developed in accordance with the ASME Code, Section XI,
Appendix G is not necessary to achieve the underlying purpose of 10 CFR
part 50, appendix G. Application of the KIM calculational
methodology of CE NPSD-683-A, Revision 6 in lieu of the calculational
methodology specified in the ASME Code, Section XI, Appendix G provides
an acceptable alternative evaluation procedure, which will continue to
meet the underlying purpose of 10 CFR part 50, appendix G. The
underlying purpose of the regulations in 10 CFR part 50, appendix G is
to provide an acceptable margin of safety against brittle failure of
[[Page 9625]]
the reactor coolant system during any condition of normal operation to
which the pressure boundary may be subjected over its service lifetime.
Based on the staff's March 16, 2001, SE regarding CE NPSD-683,
Revision 6 and the licensee's rationale to support the exemption
request, the staff agrees with the licensee's determination that an
exemption is required to approve the use of the KIM
calculational methodology of CE NPSD-683-A, Revision 6. The staff
concludes that the application of the KIM calculational
methodology of CE NPSD-683-A, Revision 6, for PVNGS, Units 1, 2, and 3
provides sufficient margin in the development of RPV P-T limit curves
such that the underlying purpose of the regulations (10 CFR part 50,
appendix G) continues to be met. Therefore, the NRC staff concludes
that the exemption requested by the licensee is justified based on the
special circumstances of 10 CFR 50.12(a)(2)(ii), ``[a]pplication of the
regulation in the particular circumstances would not serve the
underlying purpose of the rule or is not necessary to achieve the
underlying purpose of the rule.''
Based upon a consideration of the conservatism that is incorporated
into the methodologies of 10 CFR part 50, appendix G and ASME Code,
Section XI, Appendix G, the staff concludes that application of the
KIM calculational methodology of CE NPSD-683-A, Revision 6,
as described, would provide an adequate margin of safety against
brittle failure of the RPV. Therefore, the staff concludes that the
exemption is appropriate under the special circumstances of 10 CFR
50.12(a)(2)(ii), and that the application of the KIM
calculational methodology of CE NPSD-683-A, Revision 6, is acceptable
for use in the PVNGS, Units 1, 2, and 3 PTLR methodology.
4.0 Conclusion
Accordingly, the Commission has determined that, pursuant to 10 CFR
50.12(a), the exemption is authorized by law, will not present an undue
risk to the public health and safety, and is consistent with the common
defense and security. Also, special circumstances are present.
Therefore, the Commission hereby grants APS an exemption from the
requirements of 10 CFR part 50, appendix G to allow application of the
KIM calculational methodology of CE NPSD-683-A, Revision 6
in establishing the PTLR methodology for PVNGS, Units 1, 2, and 3.
Pursuant to 10 CFR 51.32, the Commission has determined that the
granting of this exemption will not have a significant effect on the
quality of the human environment (75 FR 8149; dated February 23, 2010).
This exemption is effective upon issuance.
Dated at Rockville, Maryland, this 24th day of February 2010.
For the Nuclear Regulatory Commission.
Allen G. Howe,
Acting Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. 2010-4388 Filed 3-2-10; 8:45 am]
BILLING CODE 7590-01-P