[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 (C<INF>V</INF>) and drop weight-based methodology to 
determine initial nil-ductility reference temperature 
(RT<INF>NDT</INF>) 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 RT<INF>NDT</INF> 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]]

    <bullet> 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: <a href="/cdn-cgi/l/email-protection#4d0e2c3f2221630a2c21212c2a25283f0d233f2e632a223b"><span class="__cf_email__" data-cfemail="622103100d0e4c25030e0e03050a0710220c10014c050d14">[email protected]</span></a>. For technical questions, contact 
the individual listed in the FOR FURTHER INFORMATION CONTACT section of 
this document.
    <bullet> 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 <a href="/cdn-cgi/l/email-protection#c0b0a4b2eeb2a5b3afb5b2a3a580aeb2a3eea7afb6"><span class="__cf_email__" data-cfemail="e7978395c99582948892958482a7899584c9808891">[email protected]</span></a>. The 
ADAMS accession number for each document referenced (if it is available 
in ADAMS) is provided the first time that a document is referenced.
    <bullet> 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: <a href="/cdn-cgi/l/email-protection#04456a607661652a43616b766361446a76672a636b72"><span class="__cf_email__" data-cfemail="7b3a151f091e1a553c1e14091c1e3b150918551c140d">[email protected]</span></a>.

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 RT<INF>NDT</INF> 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 (C<INF>V</INF>) and drop 
weight-based methodology to derive the initial RT<INF>NDT</INF> 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 C<INF>V</INF> 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) 
RT<INF>NDT</INF> as one that uses the procedures in ASME Code, 
Paragraph NB-2331, which requires the use of C<INF>V</INF> and drop 
weight-based test data. 10 CFR 50.61(a)(5) alternatively defines the 
method for evaluating RT<INF>NDT</INF> 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 RT<INF>NDT</INF> 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 C<INF>V</INF> 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 C<INF>V</INF> 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 RT<INF>NDT</INF> 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 
C<INF>V</INF> 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 (RT<INF>PTS</INF>) values. The licensee proposes to use the 
methodology of TR BAW-2308, Revisions 1-A and 2-A as an alternative to 
the C<INF>V</INF> and drop weight-based methodology required by 10 CFR 
50.61 for determining the initial, unirradiated properties when 
calculating RT<INF>PTS</INF>. 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 T<INF>0</INF> (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 T<INF>0</INF> values (i.e. initial unirradiated nil-ductility 
reference temperature (IRT<INF>T0</INF>) 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 C<INF>V</INF> 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 
IRT<INF>T0</INF> and [Initial Margin] [sigma]<INF>i</INF> Values for 
Linde 80 Weld Wire Heats,'' contains the NRC staff's accepted 
IRT<INF>T0</INF> and initial margin (denoted as [sigma]<INF>i</INF>) 
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 IRT<INF>T0</INF> and [sigma]<INF>i</INF> 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]RT<INF>NDT</INF> 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 IRT<INF>T0</INF> and 
[sigma]<INF>i</INF> 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 IRT<INF>T0</INF> and 
[sigma]<INF>i</INF> 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]RT<INF>NDT</INF> 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 RT<INF>NDT</INF> 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 RT<INF>NDT</INF> 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 RT<INF>NDT</INF> 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 
C<INF>V</INF> and drop weight-based methodologies to evaluate the 
initial (unirradiated) RT<INF>NDT</INF> 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


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