[Federal Register Volume 85, Number 174 (Tuesday, September 8, 2020)]
[Notices]
[Pages 55512-55514]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2020-19752]


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NUCLEAR REGULATORY COMMISSION

[Docket Nos. 50-454; 50-455; 50-456; 50-457; NRC-2020-0206]


Exelon Generation Company, LLC; Byron Station, Unit Nos. 1 and 2; 
Braidwood Station, 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 September 30, 2019, request from Exelon 
Generation Company, LLC (Exelon) from regulatory requirements to allow 
Byron Station, Unit Nos. 1 and 2, and Braidwood Station, Units 1 and 2, 
to use an alternative methodology for determining reactor coolant 
system pressure-temperature limits. The methodology is described in 
AREVA NP Topical Report BAW-2308, Revisions 1-A and 2-A, ``Initial 
RTNDT of Linde 80 Weld Materials.''

DATES: The exemption was issued on August 31, 2020.

ADDRESSES: Please refer to Docket ID NRC-2020-0206 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:
     Federal Rulemaking Website: Go to https://www.regulations.gov and search for Docket ID NRC-2020-0206. Address 
questions about Docket IDs in Regulations.gov to Jennifer Borges; 
telephone: 301-287-9127; email: [email protected]. For technical 
questions, contact the individual listed in the FOR FURTHER INFORMATION 
CONTACT section of this document.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly-available documents online in the 
ADAMS Public Documents collection at https://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``Begin Web-based ADAMS 
Search.'' For problems with ADAMS, please contact the NRC's Public 
Document Room reference staff at 1-800-397-4209, 301-415-4737, or by 
email to [email protected]. The ADAMS accession number for each 
document referenced (if it is available in ADAMS) is provided the first 
time that it is mentioned in this document.

FOR FURTHER INFORMATION CONTACT: Joel S. Wiebe, Office of Nuclear 
Reactor Regulation, U.S. Nuclear Regulatory Commission, Washington DC 
20555-0001; telephone: 301-415-6606, email: [email protected].

SUPPLEMENTARY INFORMATION: The text of the exemption is attached.

    Dated: September 1, 2020.

    For the Nuclear Regulatory Commission.
Joel S. Wiebe,
Senior Project Manager, Licensing Projects Branch III, Division of 
Operating Reactor Licensing, Office of Nuclear Reactor Regulation.

Attachment--Exemption

NUCLEAR REGULATORY COMMISSION

Docket Nos. 50-454; 50-455; 50-456; 50-457

Exelon Generation Company, LLC; Byron Station, Unit Nos. 1 and 2, and 
Braidwood Station, Units 1 and 2; Exemption

I. Background

    Exelon Generation Company, LLC (Exelon, the licensee), holds 
Renewed Facility Operating License Nos. NPF-37 and NPF-66, which 
authorize operation of the Byron Station, Unit Nos. 1 and 2 (Byron), a 
pressurized-water reactor facility, located in Ogle County, Illinois 
and Renewed Facility Operating License Nos. NPF-72 and NPF-77, which 
authorize operation of the Braidwood Station, Units 1 and 2 
(Braidwood), a pressurized-water reactor facility, located in Will 
County, Illinois. The licenses, among other things, subject the 
facilities to all rules, regulations, and orders of the U.S. Nuclear 
Regulatory Commission (NRC, the Commission) now or hereafter in effect.
    By letter dated September 30, 2019 (Agencywide Documents Access and 
Management System (ADAMS) Accession No. ML19275E307), Exelon requested 
exemptions from specific requirements of Title 10 of the Code of 
Federal Regulations (10 CFR), Part 50, Section 50.61, ``Fracture 
Toughness Requirements for Protection Against Pressurized Thermal Shock 
Events,'' and 10 CFR part 50, Appendix G, ``Fracture Toughness 
Requirements,'' for Braidwood and Byron. The requested exemptions from 
these requirements would allow use of an alternative methodology to 
determine reactor coolant system pressure-temperature limits. The new 
methodology that Exelon intends to use is described in AREVA NP Topical 
Report BAW-2308, Revisions 1-A and 2-A, ``Initial RTNDT of 
Linde 80 Weld Materials'' (BAW-2308) (ADAMS Accession No. ML032380449 
and ML081270388). BAW-2308 was approved for referencing in plant 
specific license amendments by NRC letters dated August 4, 2005 (ADAMS 
Accession No. ML052070408), and March 24, 2008 (ADAMS Accession No. 
ML080770349).

II. Request/Action

    Pursuant to 10 CFR, Part 50, Section 50.61, ``Fracture Toughness 
Requirements for Protection Against Pressurized Thermal Shock Events,'' 
and 10 CFR part 50, Appendix G, ``Fracture Toughness Requirements,'' 
the Commission's regulations establish specific fracture toughness 
requirements for nuclear power plant reactor pressure vessels (RPVs). 
In its letter dated September 30, 2019, Exelon requested exemptions 
from these requirements to allow use of an alternative methodology 
described in BAW-2308. BAW-2308 provides an alternate methodology for 
evaluating the integrity of certain RPV beltline welds, at Braidwood 
and Byron. The methodology described in BAW-2308, 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 methodology, an exemption is required 
since Appendix G to 10 CFR part 50,

[[Page 55513]]

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 Charpy 
V-notch (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 Braidwood and Byron, 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 pressurized thermal shock (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.
    Similar exemptions have been issued for Point Beach Nuclear Plant, 
Units 1 and 2 (ADAMS Accession No. ML14126A594), and Three Mile Island 
Nuclear Station, Unit 1 (ADAMS Accession No. ML13324A086).

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, as defined in 10 CFR 
50.12(a)(2). In its letter dated September 30, 2019, Exelon stated that 
the requested exemptions meet the special circumstances of 10 CFR 
50.12(a)(2)(ii), since application of the methodology in BAW-2308, in 
this particular circumstance serves the underlying purpose of the 
regulations.

A. The Exemption Is 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 Braidwood, Units 1 and 2, and Byron, Units 1 and 2, RPV Linde 80 
beltline materials and would not result in changes to operation of the 
units. 10 CFR 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. 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 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.

B. The Exemption Presents No Undue Risk to Public Health and Safety

    The NRC letter dated August 4, 2005, required licensees to meet six 
conditions and limitations to use the methods of BAW-2308 Revision 1-A. 
The NRC letter dated March 24, 2008, did not add any additional 
conditions and limitations to be resolved.
    Condition (1): By its letter dated September 30, 2019, the licensee 
provided WCAP-18370-NP, ``Braidwood Units 1 and 2 Heatup and Cooldown 
Limit Curves for Normal Operation,'' and WCAP-18371-NP, ``Byron Units 1 
and 2 Heatup and Cooldown Limit Curves for Normal Operation.'' Appendix 
G of both WCAP reports discuss the applicability of BAW-2308 to 
Braidwood and Byron Linde 80 nozzle-to-shell welds. The licensee 
compared the weld material properties of its Linde 80 nozzle-to-shell 
welds to the Linde 80 welds evaluated in BAW-2308. The licensee 
determined that the specific heats relevant to the Braidwood and Byron 
Unit 1 and 2 Linde 80 nozzle-to-shell welds were not analyzed, 
therefore, the generic ``all heats'' IRTTo and 
[sigma]1 values were used. The NRC staff reviewed the weld 
material properties of the licensee welds to those in BAW-2308 and 
confirmed that the use of the generic values was appropriate. 
Therefore, the staff determined that the licensee meets Condition (1).
    Condition (2): Section 7 in both WCAP reports discuss its 
evaluation using RG 1.99, Revision 2 method to determine the shift in 
the initial properties. Section 5 of both WCAP reports provide the 
licensee's calculation of the chemistry factors, with Tables 5-4 and 5-
5 of both reports providing the summary of chemistry factors. The NRC 
staff reviewed the chemistry factors and confirmed that the licensee 
used values greater than 167 [deg]F. The licensee provided its 
calculated adjusted reference temperature (ART) results in Tables 7-5 
and 7-8 for the extended beltline materials, including the calculated 
[Delta]RTNDT. The staff conducted confirmatory calculations 
and verified the licensee's calculated values using RG 1.99, Revision 2 
and the chemistry factors. Therefore, the staff determined that the 
licensee meets Condition (2).
    Condition (3): Tables 7-5 and 7-8 in both WCAP reports also 
provides the [sigma]I and [sigma][Delta] values 
used to calculate the ART for the extended beltline materials. The NRC 
staff confirmed that the licensee used the [sigma]I value 
from Table 3 of the NRC letter dated August 4, 2005, and 
[sigma][Delta] value of 28 [deg]F for the Linde 80 nozzle-
to-shell welds. Therefore, the NRC staff determined that the licensee 
meets Condition (3).
    Condition (4): In its letter dated September 30, 2019, the licensee 
requested an exemption, per 10 CFR 50.12 and 10 CFR 50.60(b), from the 
requirements of Appendix G to 10 CFR part 50 and 10 CFR 50.61 in 
Attachment 4 of the September 30, 2019, submittal. As part of its 
exemption request, the licensee submitted information which 
demonstrates the values the licensee proposes to use for 
[Delta]RTNDT and the margin term for each Linde 80 weld in 
its RPV through the end of its facility's current operating license. 
The exemption is addressed herein. Therefore, the NRC staff determined 
that the licensee meets Condition (4).
    Conditions (5) and (6) were resolved in BAW-2308, Revision 2, as 
documented in the NRC letter dated March 24, 2008.
    Based on the NRC reviews documented in its letters dated August 4, 
2005, and March 24, 2008, and conformance to the conditions and 
limitations as described above, the NRC staff concludes that the use of 
BAW-2308, Revisions 1-A and 2-A, does not increase the probability of 
occurrence or the consequences of an accident at Braidwood or Byron and 
will not create the possibility for a new or different type of accident 
that could pose a risk to public health and safety.
    Based on the above, the NRC finds that the action does not cause 
undue risk to public health and safety.

C. The Exemption Is Consistent With the Common Defense and Security

    The requested exemption is specifically concerned with RPV material 
properties and is consistent with guidance specified in the approved 
Topical Report BAW-2308. The exemption does not change any site 
security conditions or requirements. Therefore, the NRC finds that the 
action is consistent with the common defense and security.

[[Page 55514]]

D. Special Circumstances

    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. Application of ASME Code, Section Ill, paragraph NB-2331, in 
the determination of initial material properties was conservatively 
developed based on the level of knowledge existing in the early 1970's 
concerning RPV materials and the estimated effects of operation.
    Since the early 1970's, the level of knowledge concerning these 
topics has greatly expanded. This increased knowledge level permits 
relaxation of the ASME Code, Section Ill, paragraph NB-2331, 
requirements via application of BAW-2308, while maintaining the 
underlying purpose of the NRC regulations to ensure that an acceptable 
margin of safety is maintained.
    Based on the above, the NRC finds that use of BAW-2308 serves the 
underlying purpose of the regulation in protecting the integrity of the 
reactor coolant pressure boundary by ensuring that the RPV materials 
have adequate fracture toughness. The NRC staff has determined that 
BAW-2308 applies to the RPV materials at Braidwood and Byron, and that 
its use at these facilities is acceptable. The NRC therefore determines 
that the special circumstances required by 10 CFR 50.12(a)(2)(ii) are 
present at Braidwood and Byron.

E. Environmental Considerations

    The NRC's approval of the exemption to 10 CFR part 50, Appendix G, 
and 10 CFR 50.61 belongs to a category of actions that the NRC, by rule 
or regulation, has declared to be a categorical exclusion, after first 
finding that the category of actions does not individually or 
cumulatively have a significant effect on the human environment. 
Specifically, the exemption is categorically excluded from further 
environmental analysis under 10 CFR 51.22(c)(9).
    Under 10 CFR 51.22(c)(9), the granting of an exemption from the 
requirements of any regulation of chapter 10 of the Code of Federal 
Regulations (10 CFR) is a categorical exclusion provided that: (i) The 
exemption involves no significant hazards consideration; (ii) there is 
no significant change in the types or significant increase in the 
amounts of any effluents that may be released offsite; and (iii) there 
is no significant increase in individual or cumulative occupational 
radiation exposure.
    In its letter dated August 4, 2005, the NRC concluded that BAW-
2308, Revision 1, represents an acceptable methodology for establishing 
weld wire heat specific and generic IRTT0 values for Linde 
80 welds. In its letter dated March 24, 2008, the NRC concluded that 
that the slightly modified Pressurized-Water Reactor Owner's Group 
initial RTNDT methodology and the revised IRTT0 
and [sigma]1 values in BAW-2308, Revision 2, are acceptable 
for estimating the IRT0 and [sigma]1 values for 
various heats of the Linde 80 welds in future RPV integrity evaluations 
in license applications. Based on the above, the NRC staff has 
determined that the granting of the exemption request involves no 
significant hazards consideration because it does not: (1) Involve a 
significant increase in the probability or consequences of an accident 
previously evaluated; or (2) create the possibility of a new or 
different kind of accident from any accident previously evaluated; or 
(3) involve a significant reduction in a margin of safety. Further, the 
NRC staff has determined that issuance of the exemptions will not 
result in a significant change in the types or significant increase in 
the amounts of any effluents that may be released offsite, or a 
significant increase in individual or cumulative occupational radiation 
exposure.
    Therefore, pursuant to 10 CFR 51.22(b) and (c)(9), no environmental 
impact statement or environmental assessment need be prepared in 
connection with the approval of this exemption request.

IV. Conclusions

    Accordingly, the NRC 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 are present (see Special 
Circumstances above). Therefore, the NRC hereby grants Exelon 
Generation Company, LLC, exemptions for Byron and Braidwood, from 10 
CFR part 50, Appendix G, and 10 CFR 50.61 to allow the use of AREVA NP 
Topical Report BAW-2308, Revisions 1-A and 2-A, ``Initial 
RTNDT of Linde 80 Weld Materials.''

    Dated at Rockville, Maryland, this 31st day of August 2020

    For the Nuclear Regulatory Commission.

/RA/

Gregory F. Suber,

Deputy Director, Division of Operating Reactor Licensing, Office of 
Nuclear Reactor Regulation.

[FR Doc. 2020-19752 Filed 9-4-20; 8:45 am]
BILLING CODE 7590-01-P