[Federal Register Volume 80, Number 56 (Tuesday, March 24, 2015)]
[Notices]
[Pages 15634-15638]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2015-06700]
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NUCLEAR REGULATORY COMMISSION
[Docket No. 50-313; NRC-2015-0069]
Entergy Operations, Inc., Arkansas Nuclear One, Unit 1
AGENCY: Nuclear Regulatory Commission.
ACTION: Exemption; issuance.
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SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is issuing an
exemption in response to a March 20, 2014, request from Entergy
Operations, Inc. (Entergy or the licensee), from the requirements to
use Charpy V-notch (CV) and drop weight-based methodology to
determine initial nil-ductility reference temperature
(RTNDT) for use in evaluating the integrity of Linde 80 weld
materials in the reactor pressure vessel (RPV) beltline at Arkansas
Nuclear One (ANO), Unit 1. This exemption would allow the licensee to
use an alternate methodology to incorporate fracture toughness test
data to determine RTNDT values for use in the evaluation of
the RPV beltline weld material integrity in support of the development
of updated pressure-temperature limit curves.
DATES: March 24, 2015.
ADDRESSES: Please refer to Docket ID NRC-2015-0069 when contacting the
NRC about the availability of information regarding this document. You
may obtain publicly-available information related to this document
using any of the following methods:
[[Page 15635]]
Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2015-0069. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; 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 obtain 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 (if it 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: Andrea George, Office of Nuclear
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001; telephone: 301-415-1081, email: [email protected].
SUPPLEMENTARY INFORMATION:
I. Background
Entergy is the holder of renewed Facility Operating License No.
DPR-51, that authorizes operation of ANO, Unit 1. 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 ANO facility consists of two pressurized-water reactors, Units
1 and 2, located in Pope County, Arkansas.
II. Request/Action
Part 50 of title 10 of the Code of Federal Regulation (10 CFR),
appendix G, ``Fracture Toughness Requirements,'' specifies fracture
toughness requirements for ferritic materials of pressure-retaining
components of the reactor coolant pressure boundary of light water
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 to over its service lifetime. Section 50.61, ``Fracture
toughness requirements for protection against pressurized thermal shock
[PTS] events,'' provides fracture toughness requirements for protection
against PTS events. A PTS event is an event or transient in pressurized
water reactors (PWRs) causing severe overcooling (thermal shock)
concurrent with or followed by significant pressure in the reactor
vessel. Pursuant to 10 CFR 50.12, ``Specific exemptions,'' by letter
dated March 20, 2014 (ADAMS Accession No. ML14083A640), as supplemented
by letter dated June 26, 2014 (ADAMS Accession No. ML14177A302), the
licensee requested an exemption from certain requirements of 10 CFR
part 50, appendix G, and 10 CFR 50.61, to revise certain ANO, Unit 1
RPV initial (unirradiated) properties using AREVA Topical Report (TR)
BAW-2308, Revisions 1-A and 2-A, ``Initial RTNDT [nil-ductility
reference temperature] of Linde 80 Weld Materials.''
Specifically, the licensee requested an exemption from 10 CFR part
50, appendix G.II.D(i), which requires that licensees evaluate the pre-
service or unirradiated RTNDT according to the procedures in
the American Society of Mechanical Engineers (ASME) Code, Paragraph NB-
2331, ``Material for Vessels.'' The ASME Code Paragraph NB-2331
requires that licensees use Charpy V-notch (CV) and drop
weight-based methodology to derive the initial RTNDT values.
In lieu of the existing methodology described above, the licensee
requested to use the alternate methodology in TR BAW-2308, Revisions 1-
A and 2-A, to incorporate the use of fracture toughness test data for
evaluating the integrity of the ANO, Unit 1, Linde 80 weld materials in
the RPV beltline. The methodology in TR BAW-2308, Revisions 1-A and 2-
A, is based on the use of the 1997 and 2002 editions of the American
Society for Testing and Materials (ASTM) Standard Test Method E1921
(ASTM E1921), ``Standard Test Method for Determination of Reference
Temperature T0 for Ferritic Steels in the Transition Range,'' and ASME
Code Case N-629, ``Use of Fracture Toughness Test Data to Establish
Reference Temperature for Pressure Retaining Materials, Section III,
Division 1, Class 1.'' Since the licensee is proposing an alternate
method to the CV and drop weight-based test data required by
procedures in the ASME Code, Paragraph NB-2331, an exemption from
portions of 10 CFR part 50, appendix G, is required.
The licensee also requested an exemption from 10 CFR 50.61(a)(5),
which defines the method for evaluating initial (unirradiated)
RTNDT as one that uses the procedures in ASME Code,
Paragraph NB-2331, which requires the use of CV and drop
weight-based test data. 10 CFR 50.61(a)(5) alternatively defines the
method for evaluating RTNDT as a method other than that of
ASME Code, Paragraph NB-2331 approved by the Director, Office of
Nuclear Reactor Regulation (NRR). The licensee proposes to use the
alternate methodology described above, in AREVA TR BAW-2308,''
Revisions 1-A and 2-A, to determine the initial RTNDT values
for the Linde 80 weld materials present in the ANO, Unit 1, RPV
beltline region, which is not the procedure in ASME Code, Paragraph NB-
2331 or an alternative method approved by the Director of NRR.
Therefore, an exemption from 10 CFR 50.61(a)(5) is required.
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 and security; and
(2) when special circumstances are present. Under 10 CFR
50.12(a)(2)(ii), 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. Authorized by Law
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. Moreover, Section 50.60(b) of 10 CFR part 50 specifically
allows the use of alternative methods for determining the initial
material properties to 10 CFR part 50, appendix G, or portions thereof,
when an exemption is granted by the Commission under 10 CFR 50.12.
Because the regulations contemplate exemptions, granting the licensee's
proposed exemption will not result in a violation of the Atomic Energy
Act of 1954, as amended, or the NRC's regulations. Finally, this
exemption would allow the licensee to make use of fracture toughness
test data for evaluating the integrity of the ANO, Unit 1 RPV Linde 80
beltline weld materials, and would not result in changes to the
operation of the plant. Therefore, the exemption is authorized by law.
[[Page 15636]]
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 conditions 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 test data because
of the reference to the ASME Code, Section III, Paragraph NB-2331. The
licensee proposes to replace the use of existing CV and drop
weight-based methodology with an alternate methodology that uses
fracture toughness test data to demonstrate compliance with appendix G
to 10 CFR part 50. The alternate method, described in AREVA TR BAW-
2308, Revisions 1-A and 2-A, utilizes fracture toughness data to
determine the initial RTNDT of the Linde 80 weld materials
present in the ANO, Unit 1 RPV beltline.
The NRC staff has concluded that the requested exemption to
Appendix G to 10 CFR part 50 is justified because the licensee will
utilize 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 ANO,
Unit 1 RPV 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 the licensee under 10 CFR part
50, appendix G.II.D(i) 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 portions of 10 CFR 50.61 to use a
methodology for the determination of adjusted/indexing PTS reference
temperature (RTPTS) values. The licensee proposes to use the
methodology of TR BAW-2308, Revisions 1-A and 2-A as an alternative to
the CV and drop weight-based methodology required by 10 CFR
50.61 for determining the initial, unirradiated properties when
calculating RTPTS. The NRC has concluded that the exemption
is justified because the licensee will utilize the methodology
specified in the NRC staff's SEs regarding TR BAW-2308, Revisions 1-A
and 2-A.
In TR BAW-2308, Revision 1-A, the Babcock and Wilcox Owners Group
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 defining the initial material properties of Linde 80
welds based on T0 (initial temperature). The NRC staff
evaluated this methodology for determining Linde 80 weld initial
material properties and uncertainty in those properties, as well as the
overall method for combining initial material property measurements
based on T0 values (i.e. initial unirradiated nil-ductility
reference temperature (IRTT0) 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 defined margin term to
account for uncertainties in the NRC staff's SE for TR BAW-2308,
Revision 1-A. In the same NRC staff SE., Table 3, ``NRC Staff-Accepted
IRTT0 and [Initial Margin] [sigma]i Values for
Linde 80 Weld Wire Heats,'' contains the NRC staff's accepted
IRTT0 and initial margin (denoted as [sigma]i)
for specific Linde 80 weld wire heat numbers.
In accordance with the limitations and conditions outlined in the
NRC staff's SE for TR BAW-2308, Revision 1-A, for utilizing the values
in Table 3: The licensee has (1) utilized the appropriate NRC staff-
accepted IRTT0 and [sigma]i values for applicable
Linde 80 weld wire heat numbers; (2) applied a minimum chemistry factor
of 167 degrees Fahrenheit ([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); (3) applied
a value of 28 [deg]F for [sigma][Delta] (i.e., shift margin) in the
margin term; and (4) submitted values for [Delta]RTNDT and
the margin term for each Linde 80 weld in the RPV though the end of the
current operating license. Additionally, the NRC's SE for TR 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 TR BAW-2308, Revision 2-A and to the extent specified under Section
4.0, ``Limitations and Conditions,'' of the SE. Incidentally, although
Section 4.0 of the NRC staff SE states ``Future plant-specific
applications for RPVs containing weld heat 72105, and weld heat 299L44,
of Linde 80 must use the revised IRTT0 and
[sigma]i values in TR BAW-2308, Revision 2,'' the NRC notes
that neither of these weld heats is used at ANO, Unit 1. Therefore,
this condition does not apply to ANO, Unit 1.
During review of the licensee's exemption request, the NRC staff
noted that additional information was required in order to complete its
review regarding the chemistry factors used by the licensee for
calculating [Delta]RTNDT values. The NRC staff requested
this additional information via letter dated June 4, 2014 (ADAMS
Accession No. ML14148A382). In the licensee's supplement dated June 26,
2014, the licensee provided the chemistry factors in Table 1, ``10 CFR
50.61 Chemistry Factors for the ANO-1 RV [Reactor Vessel] Materials.''
The NRC staff confirmed that the chemistry factors used by the licensee
in calculating the RTNDT values were determined using the
methodology of RG 1.99, Revision 2, and that 167 [deg]F is the minimum
chemistry factor for Linde 80 materials.
The use of the methodology in TR BAW-2308, Revisions 1-A and 2-A,
will ensure the PTS evaluation developed for the ANO, Unit 1 RPV will
continue to be based on an adequately conservative estimate of RPV
material properties and ensure that the RPV will be protected from
failure during a PTS event. Based on the evaluations above, the NRC
staff has concluded that all
[[Page 15637]]
conditions and limitations outlined in the NRC staff's SEs for TR BAW-
2308, Revisions 1-A and 2-A, have been met for ANO Unit 1.
Based on the above information, no new accident precursors are
created by allowing an exemption to the 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 information, the
consequences of postulated accidents are not increased. Therefore there
is no undue risk to public health and safety.
D. Consistent With the Common Defense and Security
The licensee requested an exemption in order to utilize an
alternative methodology from that specified in portions of 10 CFR part
50, appendix G, and 10 CFR 50.61, to allow the use of fracture
toughness test data for evaluating the integrity of the ANO, Unit 1 RPV
beltline Linde 80 weld materials. This exemption request is not related
to, and does not impact, any security issues at ANO, Unit 1. Therefore,
the NRC has determined that this exemption does not impact, and is
consistent with, the common defense and security.
E. 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 would not serve the underlying purpose of the rule or is
not necessary to achieve the underlying purpose of the rule. The
underlying purpose of 10 CFR 50.61(a)(5) and 10 CFR part 50, appendix
G.II.D(i) is to set forth fracture toughness requirements (e.g.,
initial RTNDT values) for ferritic materials of pressure-
retaining components of the reactor coolant pressure boundary of light
water nuclear power reactors, in order to provide adequate margins of
safety during any conditions of normal operation, including anticipated
operational occurrences and system hydrostatic tests, to which the
pressure boundary may be subjected over its service lifetime. 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.
Entergy's exemption request proposes an alternate methodology to
evaluate the RTNDT of Linde 80 weld materials in the RPV
beltline region at ANO, Unit 1, based on fracture toughness test data
found in AREVA TR BAW-2308, Revision 1-A and 2-A (in accordance with
ASTM Standard E1921 and ASME Code Case N-629). This proposed alternate
methodology achieves the underlying purpose of 10 CFR part 50 appendix
G.II.D(i) because it provides an adequate conservative estimate of RPV
materials properties and ensures that the pressure-retaining components
of the RPV retain adequate margins for safety during any condition of
normal operation. The alternate methodology also achieves the
underlying purpose of 10 CFR 50.61(a)(5) because it will ensure that
the PTS evaluation developed for the ANO, Unit 1 RPV will continue to
be based on an adequately conservative estimate of RPV material
properties and ensure that the RPV will be protected from failure
during a PTS event. Accordingly, the NRC has concluded that using the
procedures in the ASME Code, Paragraph NB-2331 is not necessary to
achieve the underlying purpose of 10 CFR 50.61(a)(5) and 10 CFR part 50
appendix G.II.D(i). Therefore, the special circumstances required by 10
CFR 50.12(a)(2)(ii) for the granting of an exemption exist.
F. 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 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 pre-cursors,
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 material reference temperatures
of the Linde 80 weld materials present in the
[[Page 15638]]
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, this change does not create the possibility of a new or
different kind of accident from an accident 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 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-iii)
The proposed exemption does not make any changes to the facility,
equipment at the facility, or to fuel or core design. The proposed
alternate methodology serves the same purpose as the requirements set
forth in 10 CFR 50.61 and 10 CFR part 50, appendix G. Therefore, the
NRC concludes that the exemption involves no significant change in the
types or a significant increase in the amounts of any effluents that
may be released offsite, and that there is no significant increase in
individual or cumulative public or occupational radiation exposure.
Therefore, the requirements of 10 CFR 51.22(c)(9)(ii-iii) are met.
Conclusion
Based on the above, the NRC 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 issuance of this exemption.
IV. Conclusions
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 the licensee an exemption from
10 CFR part 50, appendix G.II.D(i) and 10 CFR 50.61(a)(5) requirements,
in order to use the alternate methodology specified in AREVA TR BAW-
2308, Revisions 1-A and 2-A, in lieu of the existing requirement to use
CV and drop weight-based methodologies to evaluate the
initial (unirradiated) RTNDT of the Linde 80 weld materials
in the RPV beltline region at ANO, Unit 1.
This exemption is effective upon issuance.
Dated at Rockville, Maryland, this 16th day of March 2015.
For the Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. 2015-06700 Filed 3-23-15; 8:45 am]
BILLING CODE 7590-01-P