FirstEnergy Nuclear Operating Company, Davis-Besse Nuclear Power Station; Exemption, 80549-80551 [2010-32141]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 75, Number 245 (Wednesday, December 22, 2010)]
[Notices]
[Pages 80549-80551]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 2010-32141]


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NUCLEAR REGULATORY COMMISSION

[Docket No. 50-346; NRC-2010-0378]


FirstEnergy Nuclear Operating Company, Davis-Besse Nuclear Power 
Station; Exemption

1.0 Background

    FirstEnergy Nuclear Operating Company (FENOC, the licensee) is the 
holder of Facility Operating License No. NFP-3, which authorizes 
operation of the Davis-Besse Nuclear Power Station, Unit 1 (DBNPS). The 
license provides, among other things, that the facility is subject to 
all rules, regulations, and orders of the U.S. Nuclear Regulatory 
Commission (NRC, the Commission) now or hereafter in effect.
    The facility consists of one pressurized-water reactor located in 
Ottawa County, Ohio.

2.0 Request/Action

    Title 10 of the Code of Federal Regulations (10 CFR), Part 50, 
Appendix G requires that fracture toughness requirements for ferritic 
materials of pressure-retaining components of the reactor coolant 
pressure boundary of light-water nuclear power reactors 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; and Section 50.61 provides fracture toughness requirements 
for protection against pressurized thermal shock (PTS) events. By 
letter dated April 15, 2009, (Agencywide Documents Access and 
Management System (ADAMS) Accession No. ML091130228), as supplemented 
by letters dated December 18, 2009, (ADAMS Accession No. ML093570103) 
and October 8, 2010 (ADAMS Accession No. ML102861221), FENOC proposed 
exemptions from the requirements of 10 CFR Part 50, Appendix G and 10 
CFR 50.61, to revise certain DBNPS reactor pressure vessel (RPV) 
initial (unirradiated) properties using Framatome Advanced Nuclear 
Power Topical Report (TR) BAW-2308, Revisions 1A and 2A, ``Initial 
RTNDT of Linde 80 Weld Materials.''
    The licensee requested an exemption from Appendix G to 10 CFR Part 
50 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 DBNPS RPV circumferential 
beltline welds 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 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 of Section III, Division 1, Class 1.'' The 
exemption is required since Appendix G to 10 CFR Part 50, through 
reference to Appendix G to Section XI of the ASME Code

[[Page 80550]]

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 the DBNPS RPV circumferential 
beltline welds based on the use of the 1997 and 2002 editions of ASTM E 
1921 and ASME Code Case N-629. 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 for 
establishing the PTS reference temperature (RTPTS).

3.0 Discussion of Exemption

    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, are consistent with the common defense and security; and (2) 
when special circumstances are present. These circumstances include the 
special circumstances that allow 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 does 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 DBNPS RPV beltline welds, and would not result in 
any changes to the operation of the plant. Section 50.60(b) of 10 CFR 
Part 50 allows the use of alternatives to 10 CFR Part 50, Appendix G, 
or portions thereof, when an exemption is granted by the Commission 
under 10 CFR 50.12. In addition, Section 50.60(b) of 10 CFR Part 50 
permits different NRC-approved methods for use in determining the 
initial material properties. As stated above, 10 CFR 50.12(a) allows 
the NRC to grant exemptions from 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. 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 exemptions are justified based on the licensee 
utilizing the fracture toughness methodology specified in BAW-2308, 
Revisions 1A and 2A, within the conditions and limitations delineated 
in the NRC staff's safety evaluations (SEs), dated August 4, 2005 
(ADAMS Accession No. ML052070408) and March 24, 2008 (ADAMS Accession 
No. ML080770349). The use of the methodology specified in the NRC 
staff's SEs will ensure that pressure-temperature limits developed for 
the DBNPS 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 Appendix G to 10 CFR Part 50, and 
does not exempt the licensee from meeting any other requirement of 
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 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 SE regarding TR BAW-2308, Revisions 1-A and 2-A, 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 nil-ductility reference temperature (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 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. 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 NRC-
accepted values of initial (unirradiated) reference temperature 
(IRTTo), 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

[[Page 80551]]

staff SE. Table 3 in the staff's August 4, 2005, SE of BAW-2308, 
Revision 1, contains the NRC staff-accepted IRTTO and 
corresponding initial uncertainty term, [sigma]I, for 
specific Linde 80 weld wire heat numbers. In accordance with the 
conditions and limitations outlined in the NRC staff's August 4, 2005 
SE of TR BAW-2308, Revision 1, for utilizing the values in Table 3, the 
licensee's proposed methodology (1) utilized the appropriate NRC staff-
accepted IRTTo and [sigma]I values for Linde 80 
weld wire heat numbers; (2) applied the appropriate chemistry factors 
for temperatures greater than 167 [deg]F (the weld wire heat-specific 
chemical composition, via the methodology of RG 1.99, Revision 2, 
indicated that higher chemistry factors are applicable); (3) applied a 
value of 28 [deg]F for [sigma][Delta] 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 heat 72105, and weld heat 299L44, 
of Linde 80 welds must use the revised IRTT0 and 
[sigma]I, values in TR BAW-2308, Revision 2.'' The staff 
notes that neither of these weld heats is used at DBNPS. Therefore, all 
conditions and limitations outlined in the NRC staff SEs for TR BAW-
2308, Revisions 1-A and 2-A, have been met for DBNPS.
    The use of the methodology in TR BAW-2308, Revision 1, will ensure 
the PTS evaluation developed for the DBNPS 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. 
Also, when additional fracture toughness data relevant to the 
evaluation of the DBNPS RPV welds is acquired as part of the 
surveillance program, this data must be incorporated into the 
evaluation of the DBNPS 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, there is no undue 
risk to public health and safety. On February 3, 2010, a new rule, 10 
CFR 50.61a, ``Alternate Fracture Toughness Requirements for Protection 
Against PTS Events,'' became effective. The NRC staff reviewed this new 
rule against the licensee's exemption request and determined that there 
is no effect on the exemption request. The new rule does not modify the 
requirements from which the licensee has sought an exemption, and the 
alternative provided by the new rule does not address the scope of 
issues associated with both 10 CFR 50.61 and 10 CFR Part 50, Appendix G 
that the requested exemption does.

Consistent With Common Defense and Security

    The proposed exemption would allow the licensee to use an alternate 
methodology to allow the use of fracture toughness test data for 
evaluating the integrity of the DBNPS RPV beltline 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 reactor vessel 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 Conclusion

    The staff has reviewed the licensee's submittals and concludes that 
the licensee has provided adequate justification for its request for an 
exemption from certain 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 DBNPS RPV beltline 
welds.
    Accordingly, the Commission has determined that pursuant to 10 CFR 
50.12, ``Specific exemptions,'' an exemption from certain requirements 
of Appendix G to 10 CFR Part 50 and 10 CFR 50.61 is authorized by law 
and will not endanger life or property or the common defense and 
security, and is otherwise in the public interest.
    Pursuant to 10 CFR 51.32, ``Finding of no significant impact,'' the 
Commission has previously determined that the granting of this 
exemption will not have a significant effect on the quality of the 
human environment (75 FR 76498).
    This exemption is effective upon issuance.

    Dated at Rockville, Maryland, this 14th day of December 2010.

    For the Nuclear Regulatory Commission.
Joseph G. Giitter,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2010-32141 Filed 12-21-10; 8:45 am]
BILLING CODE 7590-01-P