[Federal Register Volume 78, Number 247 (Tuesday, December 24, 2013)]
[Notices]
[Pages 77726-77729]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2013-30545]


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NUCLEAR REGULATORY COMMISSION

[Docket No. 50-289; NRC-2013-0274]


Exelon Generation Company, LLC Three Mile Island Nuclear Station, 
Unit 1

AGENCY: Nuclear Regulatory Commission.

ACTION: Exemption.

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SUMMARY: Exelon Generation Company, LLC (Exelon, the licensee) is the 
holder of Renewed Facility Operating License No. DPR-50, which 
authorizes operation of the Three Mile Island Nuclear Station, Unit 1 
(TMI-1). The license provides, among other things, that the facility is 
subject to all rules, regulations, and orders of the Nuclear Regulatory 
Commission (NRC) now or hereafter in effect.

ADDRESSES: Please refer to Docket ID NRC-2013-0274 when contacting the 
NRC about the availability of information regarding this document. You 
may access publicly-available information related to this action by the 
following methods:
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2013-0274. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-287-
3422; email: [email protected].
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may access publicly available documents online in the NRC 
Library 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.

SUPPLEMENTARY INFORMATION: 

1.0 Background

    Exelon Generation Company, LLC (Exelon, the licensee) is the holder 
of Renewed Facility Operating License No. DPR-50, which authorizes 
operation of the Three Mile Island Nuclear Station, Unit 1 (TMI-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 facility consists of a single pressurized-water reactor located 
in Dauphin County, Pennsylvania.

2.0 Request/Action

    Part 50, Appendix G of Title 10 of the Code of Federal Regulations 
(10 CFR),

[[Page 77727]]

``Fracture Toughness Requirements,'' specifies 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. Section 50.61, ``Fracture 
toughness requirements for protection against pressurized thermal shock 
[PTS] events,'' provides fracture toughness requirements for protection 
against PTS events. By letter dated December 14, 2012, (ADAMS) 
Accession No. ML12353A319), as supplemented by letters dated January 
31, 2013, and August 13, 2013, (ADAMS Accession Nos. ML13032A312 and 
ML13232A214, respectively), Exelon proposed exemptions from portions of 
the requirements of 10 CFR Part 50, Appendix G and 10 CFR 50.61, to 
revise certain TMI-1 reactor pressure vessel (RPV) initial 
(unirradiated) properties using AREVA Non-Proprietary Topical Report 
(TR) BAW-2308, Revisions 1A and 2A, ``Initial RTNDT [nil-
ductility reference temperature] of Linde 80 Weld Materials.''
    The licensee requested an exemption from portions of 10 CFR Part 
50, Appendix G, to replace the required use of the existing Charpy V-
notch (Cv) and drop weight-based methodology and allow the 
use of an alternate methodology to incorporate the use of fracture 
toughness test data for evaluating the integrity of the TMI-1 Linde 80 
weld materials in the RPV beltline. This request for exemption is based 
on the use of the 1997 and 2002, editions of American Society for 
Testing and Materials (ASTM) Standard Test Method E 1921 (ASTM E 1921), 
``Standard Test Method for Determination of Reference Temperature 
T0, for Ferritic Steels in the Transition Range,'' and 
American Society for Mechanical Engineering (ASME), Boiler and Pressure 
Vessel Code (Code), Code Case N-629, ``Use of Fracture Toughness Test 
Data to Establish Reference Temperature for Pressure Retaining 
Materials, Section III, Division 1, Class 1.'' Specifically, 10 CFR 
Part 50, Appendix G(II)(D)(i), requires that the nil-ductility 
reference temperature (RTNDT) be evaluated according to the 
procedures in the ASME Code, Section III, Division 1, ``Rules for 
Construction of Nuclear Power Plant Components,'' Paragraph NB-2331, 
``Material for Vessels.'' These procedures require the use of a 
methodology based on drop weight tests (NB-2331(a)(1)) and 
Cv test data (NB-2331(a)(2)). In addition, 10 CFR Part 50, 
Appendix G,(I)(A) requires the use of methods equivalent to Appendix G 
to ASME Section XI, Division 1, ``Rules for Inservice Inspection of 
Nuclear Power Plant Components,'' which specifies the use of values 
that have been determined using Cv and drop weight tests 
described above. Therefore, an exemption from portions of 10 CFR Part 
50, Appendix G, is required.
    The licensee also requested an exemption from portions of 10 CFR 
50.61 to use an alternate methodology to allow the use of fracture 
toughness test data for evaluating the integrity of the TMI-1 RPV Linde 
80 beltline welds based on the use of the 1997 and 2002, editions of 
ASTM E 1921 and ASME Code Case N-629. Similar to the above, 10 CFR 
50.61(a)(5) requires that the initial (unirradiated) RTNDT, 
be evaluated according to the procedures in the ASME Code, Section III, 
Division 1, Paragraph NB-2331. As stated previously, these procedures 
require the use of a methodology based on drop weight tests (NB-
2331(a)(1)) and Cv test data (NB-2331(a)(2)). Therefore, the 
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 to determine the initial 
RTNDT for establishing the PTS reference temperature 
(RTPTS).

3.0 Discussion

    Pursuant to 10 CFR 50.12(a), 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) special circumstances are present. The special circumstance that 
applies to these exemptions is consistent with 10 CFR 50.12(a)(2)(ii) 
in that the application of the regulations in this circumstance is not 
necessary to achieve the underlying purpose of the rules. This special 
circumstance allows the licensee an exemption from the use of the 
Cv and drop weight-based methodology required by 10 CFR Part 
50, Appendix G and 10 CFR 50.61. These exemptions only modify 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 do not exempt the licensee from meeting any other requirement of 10 
CFR Part 50, Appendix G and 10 CFR 50.61.

Authorized by Law

    These exemptions would allow the licensee to use an alternate 
methodology to make use of fracture toughness test data for evaluating 
the integrity of the TMI-1 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 granting of the licensee's proposed 
exemptions will not result in a violation of the Atomic Energy Act of 
1954, as amended, or the Commission's regulations. Therefore, the 
exemptions are authorized by law.

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. The licensee 
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 TR BAW-2308, 
Revisions 1A and 2A, 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

[[Page 77728]]

TMI-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 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 portions of 10 CFR 50.61 to use a 
methodology for the determination of adjusted/indexing reference 
temperatures. The licensee proposes to use ASME Code Case N-629 and the 
methodology outlined in its submittal, which are based on the use of 
fracture toughness data, as an alternative to the Cv and 
drop weight-based methodology required by 10 CFR 50.61 for establishing 
the initial, unirradiated properties when calculating RTPTS 
values. The NRC staff has concluded that the exemption is justified 
based on the licensee utilizing the methodology specified in the NRC 
staff's SEs regarding TR BAW-2308, Revisions 1A and 2A, dated August 4, 
2005, and March 24, 2008, respectively. This TR established an 
alternative method for determining initial (unirradiated) material 
reference temperatures for RPV welds manufactured using Linde 80 weld 
flux (i.e., ``Linde 80 welds'') and established weld wire heat-specific 
and Linde 80 weld generic values of this reference temperature. These 
weld wire heat-specific and Linde 80 weld generic values may be used in 
lieu of the RTNDT parameter, the determination of which is 
specified by paragraph NB-2331 of Section III of the ASME Code. 
Regulations associated with the determination of RPV material 
properties involving protection of the RPV from brittle failure or 
ductile rupture include Appendix G to 10 CFR Part 50 and 10 CFR 50.61, 
the PTS rule. These regulations require that the initial (unirradiated) 
material reference temperature, RTNDT, be determined in 
accordance with the provisions of the ASME Code, and provide the 
process for determination of RTPTS, the reference 
temperature RTNDT, evaluated for the end of license neutron 
fluence.
    In TR BAW-2308, Revision 1, 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 redefining the initial (unirradiated) material properties 
of Linde 80 welds. The NRC staff evaluated this methodology for 
determining Linde 80 weld initial (unirradiated) material properties 
and uncertainty in those properties, as well as the overall method for 
combining unirradiated material property measurements based on 
T0 (initial temperature) values (i.e., initial, unirradiated 
nil-ductility reference temperature (IRTT0)), 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 SE. Table 3 in the staff's SE of BAW-
2308, Revision 1, dated August 4, 2005, contains the NRC staff-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 of TR BAW-2308, Revision 1, dated August 4, 2005, for 
utilizing the values in Table 3: (1) The licensee has utilized the 
appropriate NRC staff-accepted IRTT0 and [sigma]i 
values for applicable Linde 80 weld wire heat numbers; (2) applied 
chemistry factors greater than 167 [deg]F (the weld wire heat-specific 
chemical composition, via the methodology of RG 1.99, Revision 2, 
indicated that chemistry factors higher than 167 [deg]F are 
applicable); (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 through the end of the current 
operating license. Additionally, the NRC's SE for TR BAW-2308, Revision 
2, 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, 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 wire heat 72105, 
and weld wire heat 299L44, of Linde 80 welds must use the revised 
IRTT0 and [sigma]i, values in TR BAW-2308, 
Revision 2.'' The TMI-1 RPV beltline lower nozzle belt to upper shell 
circumferential weld contains weld heat 72105. The following TMI-1 RPV 
beltline welds contain weld heat 299L44: Lower shell longitudinal weld 
(inner diameter 37 percent), and upper shell to lower shell 
circumferential weld. The licensee used the staff-accepted 
IRTT0 and [sigma]i values for Linde 80 weld 
materials containing weld wire heats 299L44 and 72105. The NRC staff 
concludes that all conditions and limitations outlined in the NRC staff 
SEs for TR BAW-2308, Revisions 1A and 2A, have been met for TMI-1.
    The use of the methodology in TR BAW-2308, Revisions 1A and 2A, 
will ensure the PTS evaluation developed for the TMI-1 RPV will 
continue to be based on an adequately conservative estimate of RPV 
material properties and ensure the RPV will be protected from failure 
during a PTS event. The NRC staff's SEs dated August 4, 2005, and March 
24, 2008, stipulate that licensees utilize the fracture toughness 
methodology, specified in TR BAW-2308, Revisions 1A and 2A, within the 
conditions and limitations delineated in the SEs.
    Based on the above information, 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 information, the 
consequences of postulated accidents are not increased. Therefore, 
there is no undue risk to public health and safety.

Consistent With Common Defense and Security

    The proposed exemptions would allow the licensee to use alternate 
methodologies from those specified in 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 TMI-1 RPV beltline

[[Page 77729]]

welds. This change has no relation to security issues. Therefore, the 
common defense and security is not impacted by these exemptions.

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 10 CFR Part 50, Appendix G and 10 CFR 
50.61 is to protect the integrity of the reactor coolant pressure 
boundary by ensuring that each RPV material has adequate fracture 
toughness. Therefore, since the underlying purpose of 10 CFR Part 50, 
Appendix G and 10 CFR 50.61 is achieved by an alternative methodology 
for evaluating RPV material fracture toughness, the special 
circumstances required by 10 CFR 50(a)(2)(ii) for the granting of an 
exemption from portions of the requirements of 10 CFR Part 50, Appendix 
G and 10 CFR 50.61 exist.

4.0 Environmental Consideration

    The exemptions would authorize exemptions from portions of the 
requirements of 10 CFR Part 50, Appendix G and 10 CFR 50.61 to allow 
the licensee to use an alternate methodology to incorporate fracture 
toughness test data for evaluating the integrity of the TMI-1 Linde 80 
weld materials in the TMI-1 RPV beltline based on the use of the 1997 
and 2002 editions of ASTM E 1921 and ASME Code Case N-629. Using the 
standard set forth in 10 CFR 50.92 for amendments to operating 
licenses, the NRC staff determined that the subject exemptions sought 
involve use of an alternate methodology to evaluate the integrity of 
the TMI-1 RPV Linde 80 beltline materials. The NRC has determined that 
these exemptions involve no significant hazards considerations:

    (1) The proposed exemptions are limited to allowing the licensee 
to use an alternative to the Cv and drop weight-based 
methodology required by 10 CFR Part 50, Appendix G and 10 CFR 50.61 
to evaluate the integrity of the TMI-1 Linde 80 weld materials in 
the TMI-1 RPV beltline. The alternate methodology does not involve 
any physical changes to the facility and does not alter the design, 
function or operation of any plant equipment. Therefore, issuance of 
this exemption does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    (2) The proposed exemption does not make any changes to the 
facility and would not create any new accident initiators. 
Therefore, this exemption does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    (3) The proposed exemption does not alter the design, function 
or operation of any plant equipment. Therefore, this exemption does 
not involve a significant reduction in a margin of safety.

    Based on the above, the NRC has concluded that the proposed 
exemptions do not involve significant hazards considerations under the 
standards set forth in 10 CFR 50.92, and accordingly, a finding of ``no 
significant hazards consideration'' is justified.
    The NRC staff has also determined that the exemptions involve no 
significant increase in the amounts, and no significant change in the 
types, of any effluents that may be released offsite; that there is no 
significant increase in individual or cumulative occupational radiation 
exposure; that there is no significant construction impact; and there 
is no significant increase in the potential for or consequences from a 
radiological accident.
    The NRC staff has further determined that the requirements from 
which the exemptions are sought involve the factors associated with 10 
CFR 51.22(c)(25)(vi)(C)--inspection or surveillance requirements. 
Specifically, the exemptions address the methodology used to develop 
the allowable pressure and temperature criteria for determining reactor 
coolant system heatup/cooldown and inservice leak and hydrostatic 
testing in accordance with Technical Specification 3.1.2, 
``Pressurization Heatup and Cooldown Limitations.'' Therefore, the 
criteria specified in 51.22(c)(25)(vi)(C) is satisfied and, 
accordingly, the exemption meets the eligibility criteria for 
categorical exclusion set forth in 10 CFR 51.22(c)(25). Pursuant to 10 
CFR 51.22(b), no environmental impact statement or environmental 
assessment is required to be prepared in connection with the issuance 
of the exemption.

5.0 Conclusion

    Accordingly, the Commission has determined that, pursuant to 10 CFR 
50.12(a), the exemptions are authorized by law, will not present an 
undue risk to the public health and safety, and are consistent with the 
common defense and security. Also, special circumstances are present. 
Therefore, the Commission hereby grants Exelon exemptions from the 
requirements of Appendix G to 10 CFR Part 50 and 10 CFR 50.61, to allow 
an alternative methodology that is based on using fracture toughness 
test data to determine initial, unirradiated properties for evaluating 
the integrity of the TMI-1 RPV beltline welds.
    This exemption is effective upon issuance.

    Dated at Rockville, Maryland, this 13th day of December 2013.

    For The Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2013-30545 Filed 12-23-13; 8:45 am]
BILLING CODE 7590-01-P