[Federal Register Volume 79, Number 135 (Tuesday, July 15, 2014)]
[Notices]
[Pages 41312-41315]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2014-16415]
=======================================================================
-----------------------------------------------------------------------
NUCLEAR REGULATORY COMMISSION
[Docket Nos. 50-266 and 50-301; NRC-2014-0167]
NextEra Energy Point Beach, LLC; Point Beach Nuclear Plant, Units
1 and 2
AGENCY: Nuclear Regulatory Commission.
ACTION: Exemption; issuance.
-----------------------------------------------------------------------
SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is issuing an
exemption in response to a January 15, 2013, letter, as supplemented on
March 1, 2013, April 18, 2013, September 12, 2013, and March 11, 2014,
from NextEra Energy Point Beach, LLC, requesting an exemption to revise
certain reactor pressure vessel (RPV) initial nil-ductility reference
temperature (RTNDT) properties using Framatome Advanced
Nuclear Power (now AREVA Nuclear Power) Topical Report BAW-2308,
Revisions 1-A and 2-A, ``Initial RTNDT of Linde 80 Weld
Materials.''
ADDRESSES: Please refer to Docket ID NRC-2014-0167 when contacting the
NRC about the availability of information regarding this document. You
may access publicly-available information related to this document
using any of the following methods:
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2014-0167. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-287-
3422; email: [email protected]. For technical questions, contact
the individual listed in the FOR FURTHER INFORMATION CONTACT section of
this document.
NRC's Agencywide Documents Access and Management System
(ADAMS): You may access publicly available documents online in the
ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS,
please contact the NRC's Public Document Room (PDR) reference staff at
1-800-397-4209, 301-415-4737, or by email to [email protected]. The
ADAMS accession number for each document referenced in this document
(if that document is available in ADAMS) is provided the first time
that a document is referenced.
NRC's PDR: You may examine and purchase copies of public
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555
Rockville Pike, Rockville, Maryland 20852.
FOR FURTHER INFORMATION CONTACT: Terry A. Beltz, Office of Nuclear
Reactor Regulation, telephone: 301-415-3049; email:
[email protected], U.S. Nuclear Regulatory Commission, Washington DC
20555-0001.
I. Background
NextEra Energy Point Beach, LLC (NextEra or the licensee) is the
holder of renewed Facility Operating License Nos. DPR-24 and DPR-27,
which authorize operation of the Point Beach Nuclear Plant (Point
Beach), Units 1 and 2, respectively. The license provides, among other
things, that the facility is subject to all rules, regulations, and
orders of the NRC now or hereafter in effect.
The facility consists of two pressurized-water reactors located in
Manitowac County in Wisconsin.
II. Request/Action
Pursuant to Section 50.12 of Title 10 of the Code of Federal
Regulations (10 CFR), ``Specific exemptions,'' the licensee has, by
letter dated January 15, 2013 (ADAMS Accession No. ML13016A208), as
supplemented on March 1, April 18, and September 12, 2013, and March
11, 2014 (ADAMS Accession Nos. ML13063A292, ML13113A008, ML13256A064,
and ML14071A405, respectively), requested an exemption from 10 CFR
50.61, ``Fracture toughness requirements for protection against
pressurized thermal shock events,'' and Appendix G to 10 CFR Part 50,
``Fracture Toughness Requirements,'' to replace the use of the required
Charpy V-notch (CV) and drop weight-based methodology with
BAW-2308, Revisions 1-A and 2-A, an alternate methodology for
evaluating the integrity of certain RPV beltline welds, at Point Beach,
Units 1 and 2. The methodology described in BAW-2308, Revisions 1-A and
2-A, utilized fracture toughness test data based on the use of the 1997
and 2002 editions of American Society for Testing and Materials (ASTM)
Standard Test Method E 1921, ``Standard Test Method for Determination
of Reference Temperature T0, for Ferritic Steels in the
Transition Range,'' and American Society for Mechanical Engineers
Boiler and Pressure Vessel Code (ASME Code), Code Case N-629, ``Use of
Fracture Toughness Test Data to establish Reference Temperature for
Pressure Retaining materials of Section III, Division 1, Class 1.''
In order to use the BAW-2308, Revision 1-A and 2-A, methodology, an
exemption is required since Appendix G to 10 CFR part 50, through
reference to Appendix G to Section XI of the ASME Code pursuant to 10
CFR 50.55(a), requires the use of a methodology based on Cv
and drop weight data.
The licensee also requested an exemption from 10 CFR 50.61 to use
an alternate methodology to allow the use of fracture toughness test
data for evaluating the integrity of certain Point Beach, Units 1 and
2, RPV beltline welds based on the use of the 1997 and 2002 editions of
ASTM E 1921 and ASME Code Case N-629. An exemption is required since
the methodology for evaluating RPV material fracture toughness in 10
CFR 50.61 requires the use of the CV and drop weight data
for establishing the PTS reference temperature (RTPTS). This
exemption only modifies the methodology to be used by the licensee for
demonstrating compliance with the requirements of 10 CFR part 50,
Appendix G and 10 CFR 50.61, and does not exempt the licensee from
meeting any other requirement of 10 CFR part 50, Appendix G and 10 CFR
50.61.
III. 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
[[Page 41313]]
and security; and (2) when special circumstances are present. Under 10
CFR 50.12(a)(2), special circumstances include, among other things,
when application of the specific regulation in the particular
circumstance would not serve, or is not necessary to achieve, the
underlying purpose of the rule.
A. Special Circumstances
Special circumstances, in accordance with 10 CFR 50.12(a)(2)(ii),
are present whenever application of the regulation in the particular
circumstances is not necessary to achieve the underlying purpose of the
rule. The underlying purpose of Appendix G to 10 CFR part 50, and 10
CFR 50.61, is to protect the integrity of the reactor coolant pressure
boundary (RCPB) by ensuring each RPV material has adequate fracture
toughness by setting forth fracture toughness requirements for ferritic
materials of pressure-retaining components of the RCPB of light water
nuclear power reactors to provide adequate margins of safety during any
condition of normal operation, including anticipated operational
occurrences and system hydrostatic tests, to which the pressure
boundary may be subjected over its service lifetime. The particular
circumstance allowing the licensee an exemption is that the use of the
alternate methodology specified in BAW-2308, Revisions 1-A and 2-A, for
evaluating the integrity of certain RPV beltline welds at Point Beach,
Units 1 and 2, continues to achieve the underlying purpose of the
rules. Therefore, the NRC staff determined that special circumstances
as required by 10 CFR 50.12(a)(2(ii) exist for granting an exemption
from portions of the requirements of 10 CFR part 50, Appendix G and 10
CFR 50.61.
B. Authorized by Law
This exemption would allow the use of an alternate methodology to
make use of fracture toughness test data for evaluating the integrity
of the Point Beach RPV Linde 80 beltline materials, and would not
result in changes to operation of the plant. Section 50.60(b) allows
the use of proposed alternatives to the described requirements in 10
CFR part 50, Appendix G, or portions thereof, when an exemption is
granted by the Commission under 10 CFR 50.12. As stated above, 10 CFR
50.12(a) allows the NRC to grant exemptions from portions of the
requirements of 10 CFR part 50, Appendix G, and 10 CFR 50.61. The NRC
staff has determined that special circumstances exist to grant the
requested exemption, and that granting the exemption will not result in
a violation of the Atomic Energy Act of 1954, as amended, or the
Commission's regulations. Therefore, the NRC staff determined that the
exemption is authorized by law.
C. No Undue Risk to Public Health and Safety
The underlying purpose of Appendix G to 10 CFR part 50 is to set
forth fracture toughness requirements for ferritic materials of
pressure-retaining components of the reactor coolant pressure boundary
of light-water nuclear power reactors to provide adequate margins of
safety during any condition of normal operation, including anticipated
operational occurrences and system hydrostatic tests, to which the
pressure boundary may be subjected over its service lifetime. The
methodology underlying the requirements of Appendix G to 10 CFR part 50
is based on the use of CV and drop weight data because of
reference to the ASME Code, as previously described. NextEra proposes
to replace the use of the existing CV and drop weight-based
methodology by a fracture toughness-based methodology to demonstrate
compliance with Appendix G to 10 CFR part 50.
The NRC staff has concluded that the requested exemption to
Appendix G to 10 CFR part 50 is justified based on the licensee
utilizing the fracture toughness methodology specified in BAW-2308,
Revisions 1-A and 2-A, within the conditions and limitations delineated
in the NRC staff's safety evaluations (SEs) dated August 4, 2005, and
March 24, 2008 (ADAMS Accession Nos. ML052070408 and ML080770349,
respectively). The use of the methodology specified in the NRC staff's
SEs will ensure that pressure-temperature limits developed for the
Point Beach, Units 1 and 2, RPVs will continue to be based on an
adequately conservative estimate of RPV material properties and ensure
that the pressure-retaining components of the reactor coolant pressure
boundary retain adequate margins of safety during any condition of
normal operation, including anticipated operational occurrences. This
exemption only modifies the methodology to be used by NextEra for
demonstrating compliance with the requirements of 10 CFR part 50,
Appendix G(II)(D)(i) and 10 CFR part 50, Appendix G(I)(A), and does not
exempt the licensee from meeting any other requirement of Appendix G to
10 CFR part 50.
Based on the above information, no new accident precursors are
created by allowing an exemption from the use of the existing
CV and drop weight-based methodology, and the use of an
alternative fracture toughness-based methodology to demonstrate
compliance with Appendix G to 10 CFR part 50; thus, the probability of
postulated accidents is not increased. Also, based on the above
information, the consequences of postulated accidents are not
increased. Therefore, there is no undue risk to public health and
safety associated with the proposed exemption to Appendix G to 10 CFR
part 50.
The underlying purpose of 10 CFR 50.61 is to establish requirements
for evaluating the fracture toughness of RPV materials to ensure that a
licensee's RPV will be protected from failure during a PTS event. The
licensee seeks an exemption from 10 CFR 50.61 to use a methodology for
the ``determination of adjusted/indexing reference temperatures.'' The
licensee proposes to use the methodology of BAW-2308, Revision 1-A, as
an alternative to the Cv and drop weight-based methodology
required by 10 CFR 50.61 for establishing the initial properties when
calculating RTPTS values. BAW-2308, Revision 2-A, is not
applicable since Point Beach does not have welds with the specific heat
numbers referenced in BAW-2308, Revision 2-A. The NRC staff has
concluded that the exemption is justified based on the licensee
utilizing the approved methodology specified in the NRC staff's SEs
regarding BAW-2308, Revision 1-A. This topical report established an
alternative method for determining initial RTPTS values for
RPV welds manufactured using Linde 80 weld flux (i.e., ``Linde 80
welds'') and established weld wire heat-specific and generic initial
RTPTS values for the Linde 80 welds. These weld wire heat-
specific and generic values may be used in lieu of the initial
RTNDT values that were determined in accordance with
Paragraph NB-2331 of Section III of the ASME Code. Appendix G to 10 CFR
part 50 and 10 CFR 50.61 require that the initial RTNDT be
determined in accordance with the provisions of the ASME Code and
provide the process for determination of RTPTS, evaluated
for the end of license fluence.
In BAW-2308, Revision 1-A, the Babcock and Wilcox Owners Group
(B&WOG) proposed to perform fracture toughness testing based on the
application of the Master Curve evaluation procedure, which permits
data obtained from sample sets tested at different temperatures to be
combined, as the basis for redefining the initial material properties
of Linde 80 welds based on T0. NRC staff evaluated this
methodology for determining Linde 80
[[Page 41314]]
weld initial material properties and uncertainty in those properties,
as well as the overall method for combining unirradiated material
property measurements based on To values (i.e.,
IRTTo in the BAW-2308 terminology), with property shifts
from models in Regulatory Guide (RG) 1.99, Revision 2, ``Radiation
Embrittlement of Reactor Vessel Materials,'' which are based on
Cv testing and a defined margin term to account for
uncertainties in the NRC staff's August 4, 2005, SE of BAW-2308,
Revision 1-A. Table 3 in the SE contains the NRC staff-accepted
IRTTo and initial margin (denoted as [sigma]i)
for specific Linde 80 weld wire heat numbers. In accordance with the
conditions and limitations outlined in the NRC staff's SE for utilizing
the values in Table 3, the licensee has utilized the appropriate NRC
staff-accepted IRTTo and [sigma]i values for
Linde 80 weld wire heat numbers; applied a minimum chemistry factor of
167[emsp14][deg]F (values greater than 167[emsp14][deg]F were used for
certain Linde 80 weld wire heat numbers if RG 1.99, Revision 2,
indicated higher chemistry factors); applied a value of
28[emsp14][deg]F for [sigma][Delta] in the margin term; and
submitted values for [Delta]RTNDT and the margin term for
each Linde 80 weld in the RPV through the end of the 50 effective full
power years (the EFPYs for the proposed P-T limits). Additionally, the
NRC's SE for BAW-2308, Revision 2-A concludes that the revised
IRTT0 and [sigma]i values for Linde 80 weld
materials are acceptable for referencing in plant-specific licensing
applications as delineated in BAW-2308, Revision 2-A and to the extent
specified under Section 4.0, Limitations and Conditions, of the SE,
which states: ``Future plant-specific applications for RPVs containing
weld heat 72105, and weld heat 299L44, of Linde 80 welds must use the
revised IRTTo and [sigma]i, values in BAW-2308,
Revision 2.'' However, the staff notes that neither of these weld heats
is used at Point Beach, Units 1 and 2. Thus, BAW-2308, Revision 2-A, is
currently not applicable. All conditions and limitations outlined in
the NRC staff SEs for BAW-2308, Revision 1-A, have been met for Point
Beach, Units 1 and 2.
The use of the methodology in BAW-2308, Revision 1-A, will ensure
the PTS evaluation developed for the Point Beach, Units 1 and 2, RPVs
will continue to be based on an adequately conservative estimate of RPV
material properties and ensure the RPVs will be protected from failure
during a PTS event. Also, when additional fracture toughness data
relevant to the evaluation of Point Beach, Units 1 and 2, RPV welds is
acquired as part of the surveillance program, this data must be
incorporated into the evaluation of the Point Beach RPV fracture
toughness requirements.
Based on the above, no new accident precursors are created by
allowing an exemption to use an alternate methodology to comply with
the requirements of 10 CFR 50.61 in determining adjusted/indexing
reference temperatures, thus, the probability of postulated accidents
is not increased. Also, based on the above, the consequences of
postulated accidents are not increased. Therefore, the NRC staff
determined that there is no undue risk to public health and safety.
D. Consistent With the Common Defense and Security
The licensee's exemption request would allow the use of alternate
methodologies from those specified in Appendix G to 10 CFR part 50, and
10 CFR 50.61, to allow the use of fracture toughness test data for
evaluating the integrity of Point Beach, Units 1 and 2, RPV beltline
welds. This change has no effect on security issues. Therefore, the NRC
staff determined that this exemption does not impact, and thus is
consistent with, the common defense and security.
E. Environmental Considerations
The NRC staff determined that the exemption discussed herein meets
the eligibility criteria for the categorical exclusion set forth in 10
CFR 51.22(c)(9) because it is related to a requirement concerning the
installation or use of a facility component located within the
restricted area, as defined in 10 CFR part 20, and issuance of this
exemption involves: (i) No significant hazards consideration, (ii) no
significant change in the types or a significant increase in the
amounts of any effluents that may be released offsite, and (iii) no
significant increase in individual or cumulative occupational radiation
exposure. Therefore, in accordance with 10 CFR 51.22(b), no
environmental impact statement or environmental assessment need be
prepared in connection with the NRC's consideration of this exemption
request. The basis for the NRC staff's determination is discussed as
follows with an evaluation against each of the requirements in 10 CFR
51.22(c)(9)(i)-(iii).
Requirements in 10 CFR 51.22(c)(9)(i)
The NRC staff evaluated whether the exemption involves no
significant hazards consideration using the standards described in 10
CFR 50.92(c), as presented below:
1. Does the proposed exemption involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The exemption would allow the use of alternate methodologies from
those specified in Appendix G to 10 CFR part 50, and 10 CFR 50.61, to
allow the use of fracture toughness test data for evaluating the
integrity of RPV beltline welds. Use of the alternate methodology for
determining the initial, unirradiated material reference temperatures
of the Linde 80 weld materials present in the RPV beltline region will
not result in changes in operation of configuration of the facility.
The change in reactor vessel material initial properties will continue
to satisfy the intent of 10 CFR 50, Appendix G, and 10 CFR 50.61. The
change does not adversely affect accident initiators or precursors, nor
alter the design assumptions, conditions, or the manner in which the
plant is operated and maintained. The change does not alter or prevent
the ability of structures, systems or components from performing their
intended function to mitigate the consequences of an initiating event
with the assumed acceptance limits. There will be no adverse change to
normal plant operating parameters, engineered safety feature actuation
setpoints, accident mitigation capabilities, or accident analysis
assumptions or inputs. The change does not affect the source term,
containment isolation, or radiological release assumptions used in
evaluating the radiological consequences of an accident previously
evaluated. Further, the change does not increase the types of amounts
of radioactive effluent that may be released offsite, nor significantly
increase individual or cumulative occupational/public radiation
exposures.
Therefore, the proposed exemption does not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
2. Does the proposed exemption create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The exemption would allow the use of alternate methodologies from
those specified in Appendix G to 10 CFR part 50, and 10 CFR 50.61, to
allow the use of fracture toughness test data for evaluating the
integrity of RPV beltline welds. Use of the alternate methodology for
determining the initial, unirradiated
[[Page 41315]]
material reference temperatures of the Linde 80 weld materials present
in the RPV beltline region will not result in changes in operation or
configuration of the facility. The change does not impose any new or
different requirements or eliminate any existing requirements. The
change is consistent with the current safety analysis assumptions and
current plant operating practice. No new accident scenarios, transient
precursors, failure mechanisms, or limiting single failures are
introduced as a result of the proposed change. Equipment important to
safety will continue to operate as designed. The change does not result
in any event previously deemed incredible being more credible. The
change does not result in any adverse conditions or result in any
increase in the challenges to safety systems.
Therefore, the proposed exemption does not create the possibility
of a new or different kind of accident from any previously evaluated.
3. Does the proposed exemption involve a significant reduction in a
margin of safety?
Response: No.
The proposed exemption does not alter safety limits, limiting
safety system settings, or limiting conditions for operation. The
setpoints at which protective actions are initiated are not altered by
the change. There are no new or significant changes to initial
conditions contributing to accident severity or consequences. The
exemption will not otherwise affect plant protective boundaries, will
not cause a release of fission products to the public, nor will it
degrade the performance of any other structures, systems or components
important to safety.
Therefore, the proposed exemption does not involve a significant
reduction in a margin of safety.
Based on the above evaluation of the standards set forth in 10 CFR
50.92(c), the NRC staff concludes that the proposed exemption involves
no significant hazards consideration. Accordingly, the requirements of
10 CFR 51.22(c)(9)(i) are met.
Requirements in 10 CFR 51.22(c)(9)(ii)
The proposed exemption would allow use of an alternate method for
determining the initial, unirradiated material reference temperatures
of the Linde 80 weld materials present in the RPV beltline region. The
proposed change in reactor vessel material initial properties will
continue to satisfy the intent of 10 CFR part 50, Appendix G, and 10
CFR 50.61. Thus, the use of this alternate methodology will not
significantly change the types of effluents that may be released
offsite, or significantly increase the amount of effluents that may be
released offsite. Therefore, the requirements of 10 CFR 51.22(c)(9)(ii)
are met.
Requirements in 10 CFR 51.22(c)(9)(iii)
The proposed exemption would allow use of an alternate method for
determining the initial, unirradiated material reference temperatures
of the Linde 80 weld materials present in the RPV beltline region. The
proposed change in reactor vessel material initial properties will
continue to satisfy the intent of 10 CFR part 50, Appendix G, and 10
CFR 50.61. Thus, the use of this alternate methodology will not
significantly increase individual occupational radiation exposure, or
significantly increase cumulative occupational radiation exposure.
Therefore, the requirements of 10 CFR 51.22(c)(9)(iii) are met.
Conclusion
Based on the above, the NRC staff concludes that the proposed
exemption meets the eligibility criteria for the categorical exclusion
set forth in 10 CFR 51.22(c)(9). Therefore, in accordance with 10 CFR
51.22(b), no environmental impact statement or environmental assessment
need be prepared in connection with the NRC's proposed issuance of this
exemption.
IV. Conclusions
Accordingly, the Commission has determined that, pursuant to 10 CFR
50.12, 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 pursuant to 10 CFR
50.12(a)(2)(ii) are present. Therefore, the Commission hereby grants
NextEra Energy Point Beach an exemption from the requirements of
Appendix G to 10 CFR part 50 and 10 CFR 50.61, to allow an alternative
methodology as described in BAW-2308, Revisions 1-A and 2-A, that is
based on using fracture toughness test data to determine initial,
unirradiated properties for evaluating the integrity of the RPV
beltline welds at the Point Beach Nuclear Plant, Units 1 and 2.
This exemption is effective upon issuance.
Dated at Rockville, Maryland, this 30th day of June 2014.
For The Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 2014-16415 Filed 7-14-14; 8:45 am]
BILLING CODE 7590-01-P