[Federal Register Volume 79, Number 135 (Tuesday, July 15, 2014)]
[Notices]
[Pages 41312-41315]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2014-16415]


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NUCLEAR REGULATORY COMMISSION

[Docket Nos. 50-266 and 50-301; NRC-2014-0167]


NextEra Energy Point Beach, LLC; Point Beach Nuclear Plant, Units 
1 and 2

AGENCY: Nuclear Regulatory Commission.

ACTION: Exemption; issuance.

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SUMMARY: The U.S. Nuclear Regulatory Commission (NRC) is issuing an 
exemption in response to a January 15, 2013, letter, as supplemented on 
March 1, 2013, April 18, 2013, September 12, 2013, and March 11, 2014, 
from NextEra Energy Point Beach, LLC, requesting an exemption to revise 
certain reactor pressure vessel (RPV) initial nil-ductility reference 
temperature (RTNDT) properties using Framatome Advanced 
Nuclear Power (now AREVA Nuclear Power) Topical Report BAW-2308, 
Revisions 1-A and 2-A, ``Initial RTNDT of Linde 80 Weld 
Materials.''

ADDRESSES: Please refer to Docket ID NRC-2014-0167 when contacting the 
NRC about the availability of information regarding this document. You 
may access publicly-available information related to this document 
using any of the following methods:
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2014-0167. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-287-
3422; email: [email protected]. For technical questions, contact 
the individual listed in the FOR FURTHER INFORMATION CONTACT section of 
this document.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may access publicly available documents online in the 
ADAMS Public Documents collection at http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and 
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS, 
please contact the NRC's Public Document Room (PDR) reference staff at 
1-800-397-4209, 301-415-4737, or by email to [email protected]. The 
ADAMS accession number for each document referenced in this document 
(if that document is available in ADAMS) is provided the first time 
that a document is referenced.
     NRC's PDR: You may examine and purchase copies of public 
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852.

FOR FURTHER INFORMATION CONTACT: Terry A. Beltz, Office of Nuclear 
Reactor Regulation, telephone: 301-415-3049; email: 
[email protected], U.S. Nuclear Regulatory Commission, Washington DC 
20555-0001.

I. Background

    NextEra Energy Point Beach, LLC (NextEra or the licensee) is the 
holder of renewed Facility Operating License Nos. DPR-24 and DPR-27, 
which authorize operation of the Point Beach Nuclear Plant (Point 
Beach), Units 1 and 2, respectively. The license provides, among other 
things, that the facility is subject to all rules, regulations, and 
orders of the NRC now or hereafter in effect.
    The facility consists of two pressurized-water reactors located in 
Manitowac County in Wisconsin.

II. Request/Action

    Pursuant to Section 50.12 of Title 10 of the Code of Federal 
Regulations (10 CFR), ``Specific exemptions,'' the licensee has, by 
letter dated January 15, 2013 (ADAMS Accession No. ML13016A208), as 
supplemented on March 1, April 18, and September 12, 2013, and March 
11, 2014 (ADAMS Accession Nos. ML13063A292, ML13113A008, ML13256A064, 
and ML14071A405, respectively), requested an exemption from 10 CFR 
50.61, ``Fracture toughness requirements for protection against 
pressurized thermal shock events,'' and Appendix G to 10 CFR Part 50, 
``Fracture Toughness Requirements,'' to replace the use of the required 
Charpy V-notch (CV) and drop weight-based methodology with 
BAW-2308, Revisions 1-A and 2-A, an alternate methodology for 
evaluating the integrity of certain RPV beltline welds, at Point Beach, 
Units 1 and 2. The methodology described in BAW-2308, Revisions 1-A and 
2-A, utilized fracture toughness test data based on the use of the 1997 
and 2002 editions of American Society for Testing and Materials (ASTM) 
Standard Test Method E 1921, ``Standard Test Method for Determination 
of Reference Temperature T0, for Ferritic Steels in the 
Transition Range,'' and American Society for Mechanical Engineers 
Boiler and Pressure Vessel Code (ASME Code), Code Case N-629, ``Use of 
Fracture Toughness Test Data to establish Reference Temperature for 
Pressure Retaining materials of Section III, Division 1, Class 1.''
    In order to use the BAW-2308, Revision 1-A and 2-A, methodology, an 
exemption is required since Appendix G to 10 CFR part 50, through 
reference to Appendix G to Section XI of the ASME Code pursuant to 10 
CFR 50.55(a), requires the use of a methodology based on Cv 
and drop weight data.
    The licensee also requested an exemption from 10 CFR 50.61 to use 
an alternate methodology to allow the use of fracture toughness test 
data for evaluating the integrity of certain Point Beach, Units 1 and 
2, RPV beltline welds based on the use of the 1997 and 2002 editions of 
ASTM E 1921 and ASME Code Case N-629. An exemption is required since 
the methodology for evaluating RPV material fracture toughness in 10 
CFR 50.61 requires the use of the CV and drop weight data 
for establishing the PTS reference temperature (RTPTS). This 
exemption only modifies the methodology to be used by the licensee for 
demonstrating compliance with the requirements of 10 CFR part 50, 
Appendix G and 10 CFR 50.61, and does not exempt the licensee from 
meeting any other requirement of 10 CFR part 50, Appendix G and 10 CFR 
50.61.

III. Discussion

    Pursuant to 10 CFR 50.12, the Commission may, upon application by 
any interested person or upon its own initiative, grant exemptions from 
the requirements of 10 CFR part 50 when: (1) the exemptions are 
authorized by law, will not present an undue risk to public health or 
safety, and are consistent with the common defense

[[Page 41313]]

and security; and (2) when special circumstances are present. Under 10 
CFR 50.12(a)(2), special circumstances include, among other things, 
when application of the specific regulation in the particular 
circumstance would not serve, or is not necessary to achieve, the 
underlying purpose of the rule.

A. Special Circumstances

    Special circumstances, in accordance with 10 CFR 50.12(a)(2)(ii), 
are present whenever application of the regulation in the particular 
circumstances is not necessary to achieve the underlying purpose of the 
rule. The underlying purpose of Appendix G to 10 CFR part 50, and 10 
CFR 50.61, is to protect the integrity of the reactor coolant pressure 
boundary (RCPB) by ensuring each RPV material has adequate fracture 
toughness by setting forth fracture toughness requirements for ferritic 
materials of pressure-retaining components of the RCPB of light water 
nuclear power reactors to provide adequate margins of safety during any 
condition of normal operation, including anticipated operational 
occurrences and system hydrostatic tests, to which the pressure 
boundary may be subjected over its service lifetime. The particular 
circumstance allowing the licensee an exemption is that the use of the 
alternate methodology specified in BAW-2308, Revisions 1-A and 2-A, for 
evaluating the integrity of certain RPV beltline welds at Point Beach, 
Units 1 and 2, continues to achieve the underlying purpose of the 
rules. Therefore, the NRC staff determined that special circumstances 
as required by 10 CFR 50.12(a)(2(ii) exist for granting an exemption 
from portions of the requirements of 10 CFR part 50, Appendix G and 10 
CFR 50.61.

B. Authorized by Law

    This exemption would allow the use of an alternate methodology to 
make use of fracture toughness test data for evaluating the integrity 
of the Point Beach RPV Linde 80 beltline materials, and would not 
result in changes to operation of the plant. Section 50.60(b) allows 
the use of proposed alternatives to the described requirements in 10 
CFR part 50, Appendix G, or portions thereof, when an exemption is 
granted by the Commission under 10 CFR 50.12. As stated above, 10 CFR 
50.12(a) allows the NRC to grant exemptions from portions of the 
requirements of 10 CFR part 50, Appendix G, and 10 CFR 50.61. The NRC 
staff has determined that special circumstances exist to grant the 
requested exemption, and that granting the exemption will not result in 
a violation of the Atomic Energy Act of 1954, as amended, or the 
Commission's regulations. Therefore, the NRC staff determined that the 
exemption is authorized by law.

C. No Undue Risk to Public Health and Safety

    The underlying purpose of Appendix G to 10 CFR part 50 is to set 
forth fracture toughness requirements for ferritic materials of 
pressure-retaining components of the reactor coolant pressure boundary 
of light-water nuclear power reactors to provide adequate margins of 
safety during any condition of normal operation, including anticipated 
operational occurrences and system hydrostatic tests, to which the 
pressure boundary may be subjected over its service lifetime. The 
methodology underlying the requirements of Appendix G to 10 CFR part 50 
is based on the use of CV and drop weight data because of 
reference to the ASME Code, as previously described. NextEra proposes 
to replace the use of the existing CV and drop weight-based 
methodology by a fracture toughness-based methodology to demonstrate 
compliance with Appendix G to 10 CFR part 50.
    The NRC staff has concluded that the requested exemption to 
Appendix G to 10 CFR part 50 is justified based on the licensee 
utilizing the fracture toughness methodology specified in BAW-2308, 
Revisions 1-A and 2-A, within the conditions and limitations delineated 
in the NRC staff's safety evaluations (SEs) dated August 4, 2005, and 
March 24, 2008 (ADAMS Accession Nos. ML052070408 and ML080770349, 
respectively). The use of the methodology specified in the NRC staff's 
SEs will ensure that pressure-temperature limits developed for the 
Point Beach, Units 1 and 2, RPVs will continue to be based on an 
adequately conservative estimate of RPV material properties and ensure 
that the pressure-retaining components of the reactor coolant pressure 
boundary retain adequate margins of safety during any condition of 
normal operation, including anticipated operational occurrences. This 
exemption only modifies the methodology to be used by NextEra for 
demonstrating compliance with the requirements of 10 CFR part 50, 
Appendix G(II)(D)(i) and 10 CFR part 50, Appendix G(I)(A), and does not 
exempt the licensee from meeting any other requirement of Appendix G to 
10 CFR part 50.
    Based on the above information, no new accident precursors are 
created by allowing an exemption from the use of the existing 
CV and drop weight-based methodology, and the use of an 
alternative fracture toughness-based methodology to demonstrate 
compliance with Appendix G to 10 CFR part 50; thus, the probability of 
postulated accidents is not increased. Also, based on the above 
information, the consequences of postulated accidents are not 
increased. Therefore, there is no undue risk to public health and 
safety associated with the proposed exemption to Appendix G to 10 CFR 
part 50.
    The underlying purpose of 10 CFR 50.61 is to establish requirements 
for evaluating the fracture toughness of RPV materials to ensure that a 
licensee's RPV will be protected from failure during a PTS event. The 
licensee seeks an exemption from 10 CFR 50.61 to use a methodology for 
the ``determination of adjusted/indexing reference temperatures.'' The 
licensee proposes to use the methodology of BAW-2308, Revision 1-A, as 
an alternative to the Cv and drop weight-based methodology 
required by 10 CFR 50.61 for establishing the initial properties when 
calculating RTPTS values. BAW-2308, Revision 2-A, is not 
applicable since Point Beach does not have welds with the specific heat 
numbers referenced in BAW-2308, Revision 2-A. The NRC staff has 
concluded that the exemption is justified based on the licensee 
utilizing the approved methodology specified in the NRC staff's SEs 
regarding BAW-2308, Revision 1-A. This topical report established an 
alternative method for determining initial RTPTS values for 
RPV welds manufactured using Linde 80 weld flux (i.e., ``Linde 80 
welds'') and established weld wire heat-specific and generic initial 
RTPTS values for the Linde 80 welds. These weld wire heat-
specific and generic values may be used in lieu of the initial 
RTNDT values that were determined in accordance with 
Paragraph NB-2331 of Section III of the ASME Code. Appendix G to 10 CFR 
part 50 and 10 CFR 50.61 require that the initial RTNDT be 
determined in accordance with the provisions of the ASME Code and 
provide the process for determination of RTPTS, evaluated 
for the end of license fluence.
    In BAW-2308, Revision 1-A, the Babcock and Wilcox Owners Group 
(B&WOG) proposed to perform fracture toughness testing based on the 
application of the Master Curve evaluation procedure, which permits 
data obtained from sample sets tested at different temperatures to be 
combined, as the basis for redefining the initial material properties 
of Linde 80 welds based on T0. NRC staff evaluated this 
methodology for determining Linde 80

[[Page 41314]]

weld initial material properties and uncertainty in those properties, 
as well as the overall method for combining unirradiated material 
property measurements based on To values (i.e., 
IRTTo in the BAW-2308 terminology), with property shifts 
from models in Regulatory Guide (RG) 1.99, Revision 2, ``Radiation 
Embrittlement of Reactor Vessel Materials,'' which are based on 
Cv testing and a defined margin term to account for 
uncertainties in the NRC staff's August 4, 2005, SE of BAW-2308, 
Revision 1-A. Table 3 in the SE contains the NRC staff-accepted 
IRTTo and initial margin (denoted as [sigma]i) 
for specific Linde 80 weld wire heat numbers. In accordance with the 
conditions and limitations outlined in the NRC staff's SE for utilizing 
the values in Table 3, the licensee has utilized the appropriate NRC 
staff-accepted IRTTo and [sigma]i values for 
Linde 80 weld wire heat numbers; applied a minimum chemistry factor of 
167[emsp14][deg]F (values greater than 167[emsp14][deg]F were used for 
certain Linde 80 weld wire heat numbers if RG 1.99, Revision 2, 
indicated higher chemistry factors); applied a value of 
28[emsp14][deg]F for [sigma][Delta] in the margin term; and 
submitted values for [Delta]RTNDT and the margin term for 
each Linde 80 weld in the RPV through the end of the 50 effective full 
power years (the EFPYs for the proposed P-T limits). Additionally, the 
NRC's SE for BAW-2308, Revision 2-A concludes that the revised 
IRTT0 and [sigma]i values for Linde 80 weld 
materials are acceptable for referencing in plant-specific licensing 
applications as delineated in BAW-2308, Revision 2-A and to the extent 
specified under Section 4.0, Limitations and Conditions, of the SE, 
which states: ``Future plant-specific applications for RPVs containing 
weld heat 72105, and weld heat 299L44, of Linde 80 welds must use the 
revised IRTTo and [sigma]i, values in BAW-2308, 
Revision 2.'' However, the staff notes that neither of these weld heats 
is used at Point Beach, Units 1 and 2. Thus, BAW-2308, Revision 2-A, is 
currently not applicable. All conditions and limitations outlined in 
the NRC staff SEs for BAW-2308, Revision 1-A, have been met for Point 
Beach, Units 1 and 2.
    The use of the methodology in BAW-2308, Revision 1-A, will ensure 
the PTS evaluation developed for the Point Beach, Units 1 and 2, RPVs 
will continue to be based on an adequately conservative estimate of RPV 
material properties and ensure the RPVs will be protected from failure 
during a PTS event. Also, when additional fracture toughness data 
relevant to the evaluation of Point Beach, Units 1 and 2, RPV welds is 
acquired as part of the surveillance program, this data must be 
incorporated into the evaluation of the Point Beach RPV fracture 
toughness requirements.
    Based on the above, no new accident precursors are created by 
allowing an exemption to use an alternate methodology to comply with 
the requirements of 10 CFR 50.61 in determining adjusted/indexing 
reference temperatures, thus, the probability of postulated accidents 
is not increased. Also, based on the above, the consequences of 
postulated accidents are not increased. Therefore, the NRC staff 
determined that there is no undue risk to public health and safety.

D. Consistent With the Common Defense and Security

    The licensee's exemption request would allow the use of alternate 
methodologies from those specified in Appendix G to 10 CFR part 50, and 
10 CFR 50.61, to allow the use of fracture toughness test data for 
evaluating the integrity of Point Beach, Units 1 and 2, RPV beltline 
welds. This change has no effect on security issues. Therefore, the NRC 
staff determined that this exemption does not impact, and thus is 
consistent with, the common defense and security.

E. Environmental Considerations

    The NRC staff determined that the exemption discussed herein meets 
the eligibility criteria for the categorical exclusion set forth in 10 
CFR 51.22(c)(9) because it is related to a requirement concerning the 
installation or use of a facility component located within the 
restricted area, as defined in 10 CFR part 20, and issuance of this 
exemption involves: (i) No significant hazards consideration, (ii) no 
significant change in the types or a significant increase in the 
amounts of any effluents that may be released offsite, and (iii) no 
significant increase in individual or cumulative occupational radiation 
exposure. Therefore, in accordance with 10 CFR 51.22(b), no 
environmental impact statement or environmental assessment need be 
prepared in connection with the NRC's consideration of this exemption 
request. The basis for the NRC staff's determination is discussed as 
follows with an evaluation against each of the requirements in 10 CFR 
51.22(c)(9)(i)-(iii).

Requirements in 10 CFR 51.22(c)(9)(i)

    The NRC staff evaluated whether the exemption involves no 
significant hazards consideration using the standards described in 10 
CFR 50.92(c), as presented below:
1. Does the proposed exemption involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The exemption would allow the use of alternate methodologies from 
those specified in Appendix G to 10 CFR part 50, and 10 CFR 50.61, to 
allow the use of fracture toughness test data for evaluating the 
integrity of RPV beltline welds. Use of the alternate methodology for 
determining the initial, unirradiated material reference temperatures 
of the Linde 80 weld materials present in the RPV beltline region will 
not result in changes in operation of configuration of the facility. 
The change in reactor vessel material initial properties will continue 
to satisfy the intent of 10 CFR 50, Appendix G, and 10 CFR 50.61. The 
change does not adversely affect accident initiators or precursors, nor 
alter the design assumptions, conditions, or the manner in which the 
plant is operated and maintained. The change does not alter or prevent 
the ability of structures, systems or components from performing their 
intended function to mitigate the consequences of an initiating event 
with the assumed acceptance limits. There will be no adverse change to 
normal plant operating parameters, engineered safety feature actuation 
setpoints, accident mitigation capabilities, or accident analysis 
assumptions or inputs. The change does not affect the source term, 
containment isolation, or radiological release assumptions used in 
evaluating the radiological consequences of an accident previously 
evaluated. Further, the change does not increase the types of amounts 
of radioactive effluent that may be released offsite, nor significantly 
increase individual or cumulative occupational/public radiation 
exposures.
    Therefore, the proposed exemption does not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
2. Does the proposed exemption create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The exemption would allow the use of alternate methodologies from 
those specified in Appendix G to 10 CFR part 50, and 10 CFR 50.61, to 
allow the use of fracture toughness test data for evaluating the 
integrity of RPV beltline welds. Use of the alternate methodology for 
determining the initial, unirradiated

[[Page 41315]]

material reference temperatures of the Linde 80 weld materials present 
in the RPV beltline region will not result in changes in operation or 
configuration of the facility. The change does not impose any new or 
different requirements or eliminate any existing requirements. The 
change is consistent with the current safety analysis assumptions and 
current plant operating practice. No new accident scenarios, transient 
precursors, failure mechanisms, or limiting single failures are 
introduced as a result of the proposed change. Equipment important to 
safety will continue to operate as designed. The change does not result 
in any event previously deemed incredible being more credible. The 
change does not result in any adverse conditions or result in any 
increase in the challenges to safety systems.
    Therefore, the proposed exemption does not create the possibility 
of a new or different kind of accident from any previously evaluated.
3. Does the proposed exemption involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed exemption does not alter safety limits, limiting 
safety system settings, or limiting conditions for operation. The 
setpoints at which protective actions are initiated are not altered by 
the change. There are no new or significant changes to initial 
conditions contributing to accident severity or consequences. The 
exemption will not otherwise affect plant protective boundaries, will 
not cause a release of fission products to the public, nor will it 
degrade the performance of any other structures, systems or components 
important to safety.
    Therefore, the proposed exemption does not involve a significant 
reduction in a margin of safety.
    Based on the above evaluation of the standards set forth in 10 CFR 
50.92(c), the NRC staff concludes that the proposed exemption involves 
no significant hazards consideration. Accordingly, the requirements of 
10 CFR 51.22(c)(9)(i) are met.

Requirements in 10 CFR 51.22(c)(9)(ii)

    The proposed exemption would allow use of an alternate method for 
determining the initial, unirradiated material reference temperatures 
of the Linde 80 weld materials present in the RPV beltline region. The 
proposed change in reactor vessel material initial properties will 
continue to satisfy the intent of 10 CFR part 50, Appendix G, and 10 
CFR 50.61. Thus, the use of this alternate methodology will not 
significantly change the types of effluents that may be released 
offsite, or significantly increase the amount of effluents that may be 
released offsite. Therefore, the requirements of 10 CFR 51.22(c)(9)(ii) 
are met.

Requirements in 10 CFR 51.22(c)(9)(iii)

    The proposed exemption would allow use of an alternate method for 
determining the initial, unirradiated material reference temperatures 
of the Linde 80 weld materials present in the RPV beltline region. The 
proposed change in reactor vessel material initial properties will 
continue to satisfy the intent of 10 CFR part 50, Appendix G, and 10 
CFR 50.61. Thus, the use of this alternate methodology will not 
significantly increase individual occupational radiation exposure, or 
significantly increase cumulative occupational radiation exposure. 
Therefore, the requirements of 10 CFR 51.22(c)(9)(iii) are met.

Conclusion

    Based on the above, the NRC staff concludes that the proposed 
exemption meets the eligibility criteria for the categorical exclusion 
set forth in 10 CFR 51.22(c)(9). Therefore, in accordance with 10 CFR 
51.22(b), no environmental impact statement or environmental assessment 
need be prepared in connection with the NRC's proposed issuance of this 
exemption.

IV. Conclusions

    Accordingly, the Commission has determined that, pursuant to 10 CFR 
50.12, the exemption is authorized by law, will not present an undue 
risk to the public health and safety, and is consistent with the common 
defense and security. Also, special circumstances pursuant to 10 CFR 
50.12(a)(2)(ii) are present. Therefore, the Commission hereby grants 
NextEra Energy Point Beach an exemption from the requirements of 
Appendix G to 10 CFR part 50 and 10 CFR 50.61, to allow an alternative 
methodology as described in BAW-2308, Revisions 1-A and 2-A, that is 
based on using fracture toughness test data to determine initial, 
unirradiated properties for evaluating the integrity of the RPV 
beltline welds at the Point Beach Nuclear Plant, Units 1 and 2.
    This exemption is effective upon issuance.

    Dated at Rockville, Maryland, this 30th day of June 2014.

    For The Nuclear Regulatory Commission.
Michele G. Evans,
Director, Division of Operating Reactor Licensing, Office of Nuclear 
Reactor Regulation.
[FR Doc. 2014-16415 Filed 7-14-14; 8:45 am]
BILLING CODE 7590-01-P