25554 Federal Register / Vol. 87, No. 83 / Friday, April 29, 2022 / Notices 

participants separately. Therefore, 

applicants and other participants (or 

their counsel or representative) must 

apply for and receive a digital ID 

certificate before adjudicatory 

documents are filed so that they can 

obtain access to the documents via the 

E-Filing system. 

A person filing electronically using 

the NRC’s adjudicatory E-Filing system 

may seek assistance by contacting the 

NRC’s Electronic Filing Help Desk 

through the ‘‘Contact Us’’ link located 

on the NRC’s public website at http://

www.nrc.gov/site-help/esubmittals.html, by email to 

MSHD.Resource@nrc.gov, or by a tollfree call at 1–866–672–7640. The NRC 

Electronic Filing Help Desk is available 

between 9 a.m. and 6 p.m., Eastern 

Time, Monday through Friday, 

excluding government holidays. 

Participants who believe that they 

have a good cause for not submitting 

documents electronically must file an 

exemption request, in accordance with 

10 CFR 2.302(g), with their initial paper 

filing requesting authorization to 

continue to submit documents in paper 

format. Such filings must be submitted 

by: (1) First class mail addressed to the 

Office of the Secretary of the 

Commission, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555– 

0001, Attention: Rulemaking and 

Adjudications Staff; or (2) courier, 

express mail, or expedited delivery 

service to the Office of the Secretary, 

Sixteenth Floor, One White Flint North, 

11555 Rockville Pike, Rockville, 

Maryland, 20852, Attention: 

Rulemaking and Adjudications Staff. 

Participants filing a document in this 

manner are responsible for serving the 

document on all other participants. 

Filing is considered complete by firstclass mail as of the time of deposit in 

the mail, or by courier, express mail, or 

expedited delivery service upon 

depositing the document with the 

provider of the service. A presiding 

officer, having granted an exemption 

request from using E-Filing, may require 

a participant or party to use E-Filing if 

the presiding officer subsequently 

determines that the reason for granting 

the exemption from use of E-Filing no 

longer exists. 

Documents submitted in adjudicatory 

proceedings will appear in NRC’s 

electronic hearing docket, which is 

available to the public at https://

adams.nrc.gov/ehd, unless excluded 

pursuant to an order of the Commission 

or the presiding officer. Participants are 

requested not to include personal 

privacy information, such as social 

security numbers, home addresses, or 

home phone numbers in their filings, 

unless an NRC regulation or other law 

requires submission of such 

information. With respect to 

copyrighted works, except for limited 

excerpts that serve the purpose of the 

adjudicatory filings and would 

constitute a Fair Use application, 

participants are requested not to include 

copyrighted materials in their 

submission. 

The Commission will issue a notice or 

order granting or denying a hearing 

request or intervention petition, 

designating the issues for any hearing 

that will be held and designating the 

Presiding Officer. A notice granting a 

hearing will be published in the Federal 

Register and served on the parties to the 

hearing. 

If a person (other than VA) requests a 

hearing, that person shall set forth with 

particularity the manner in which his 

interest is adversely affected by this 

Confirmatory Order and shall address 

the criteria set forth in 10 CFR 2.309(d) 

and (f). 

If a hearing is requested by a person 

whose interest is adversely affected, the 

Commission will issue an order 

designating the time and place of any 

hearings. If a hearing is held, the issue 

to be considered at such hearing shall be 

whether this Confirmatory Order should 

be sustained. 

In the absence of any request for 

hearing, or written approval of an 

extension of time in which to request a 

hearing, the provisions specified in 

Section V above shall be final 30 days 

from the date of this Confirmatory Order 

without further order or proceedings. If 

an extension of time for requesting a 

hearing has been approved, the 

provisions specified in Section V shall 

be final when the extension expires if a 

hearing request has not been received. 

For the Nuclear Regulatory Commission. 

Dated this 21st day of April 2022. 

/RA/ 

John B. Giessner, 

Regional Administrator, NRC Region III. 

[FR Doc. 2022–09168 Filed 4–28–22; 8:45 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[Docket Nos. 72–1041, 50–498, and 50–499; 

NRC–2022–0099] 

South Texas Project Nuclear Operating 

Company; South Texas Project Electric 

Generating Station Units 1 and 2; 

Independent Spent Fuel Storage 

Installation 

AGENCY: Nuclear Regulatory 

Commission. 

ACTION: Exemption; issuance. 

SUMMARY: The U.S. Nuclear Regulatory 

Commission (NRC) is issuing an 

exemption in response to a request 

submitted by South Texas Project 

Nuclear Operating Company (STPNOC) 

on March 11, 2022, from meeting certain 

NRC regulatory requirements for one 

multipurpose canister (MPC), Serial 

Number 248 (MPC 248), in use at the 

South Texas Project Electric Generating 

Station, Units 1 and 2 (STPEGS). This 

exemption permits STPNOC to continue 

using MPC 248 to store spent fuel for 

the service life of the canister, including 

transferring the MPC to a HI–STORM 

FW overpack, without volumetric 

examination data from radiographic 

testing for a 1-inch section of the 

repaired weld seam joining the 

baseplate to the canister shell. 

DATES: This exemption was issued on 

April 25, 2022. 

ADDRESSES: Please refer to Docket ID 

NRC–2022–0099 when contacting the 

NRC about the availability of 

information regarding this action. You 

may obtain publicly available 

information related to this action using 

any of the following methods: 

• Federal Rulemaking Website: Go to 

https://www.regulations.gov and search 

for Docket ID NRC–2022–0099. Address 

questions about Docket IDs in 

Regulations.gov to Stacy Schumann; 

telephone: 301–415–0624; email: 

Stacy.Schumann@nrc.gov. 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 (PDR) 

reference staff at 1–800–397–4209, 301– 

415–4737, or by email to 

PDR.Resource@nrc.gov. For the 

convenience of the reader, instructions 

about obtaining materials referenced in 

this document are provided in the 

‘‘Availability of Documents’’ section. 

• NRC’s PDR: You may examine and 

purchase copies of public documents, 

by appointment, at the NRC’s PDR, 

Room P1 B35, One White Flint North, 

11555 Rockville Pike, Rockville, 

Maryland 20852. To make an 

appointment to visit the PDR, please 

send an email to PDR.Resource@nrc.gov 

or call 1–800–397–4209 or 301–415– 

4737, between 8:00 a.m. and 4:00 p.m. 


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(ET), Monday through Friday, except 

Federal holidays. 

FOR FURTHER INFORMATION CONTACT: 

Donald Habib, Office of Nuclear 

Material Safety and Safeguards, U.S. 

Nuclear Regulatory Commission, 

Washington, DC 20555–0001; telephone: 

301–415–1035; email: Donald.Habib@

nrc.gov. 

SUPPLEMENTARY INFORMATION: 

I. Background 

South Texas Project Nuclear 

Operating Company (STPNOC or the 

licensee) is the holder of Facility 

Operating License Nos. NPF–76 and 

NPF–80, which authorize operation of 

the STPEGS, respectively, in Matagorda 

County, Texas, pursuant to part 50 of 

title 10 of the Code of Federal 

Regulations (10 CFR), ‘‘Domestic 

Licensing of Production and Utilization 

Facilities.’’ 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. 

Under 10 CFR part 72, subpart K, 

‘‘General License for Storage of Spent 

Fuel at Power Reactor Sites,’’ a general 

license is issued for the storage of spent 

fuel in an independent spent fuel 

storage installation (ISFSI) at power 

reactor sites to persons authorized to 

possess or operate nuclear power 

reactors under 10 CFR part 50. The 

licensee is authorized to operate a 

nuclear power reactor under 10 CFR 

part 50 and, accordingly, holds a 10 CFR 

part 72 general license for storage of 

spent fuel at the STPEGS ISFSI. Under 

10 CFR 72.212(a)(2), (b)(3), (b)(5)(i), 

(b)(11) and 72.214, a general licensee 

may store spent fuel in a cask, so long 

as it is one of the approved casks listed 

in 10 CFR 72.214 and the general 

licensee conforms to the terms, 

conditions, and specifications of the 

relevant certificate of compliance (CoC) 

or amended CoC. Accordingly, under 

the terms of the general license, the 

STPNOC stores spent fuel at its ISFSI 

using the HI–STORM FW MPC–37 

Storage System in accordance with CoC 

No. 1032, Amendment No. 2. As part of 

the MPC storage system, the MPC (of 

which the weld seam joining the 

baseplate to the shell is an integral part) 

ensures the functions of criticality 

safety, confinement boundary, 

shielding, structural support, and heat 

transfer. 

II. Request/Action 

In a letter dated March 11, 2022, the 

licensee requested an exemption from 

the requirements of 10 CFR 72.154(b) as 

well as 10 CFR 72.212(a)(2), (b)(3), 

(b)(5)(i), and (b)(11). Paragraph 

72.154(b) requires the licensee to have 

available documentary evidence that 

material and equipment conform to the 

procurement specifications prior to 

installation or use of the material and 

equipment and to retain or have 

available this documentary evidence for 

the life of the ISFSI or spent fuel cask. 

Paragraph 72.212(a)(2) limits a general 

license to storage of spent fuel in casks 

approved under the provisions of 10 

CFR part 72. Paragraph 72.212(b)(3) 

requires the general licensee to ensure 

that each cask it uses conforms to the 

terms, conditions, and specifications of 

a CoC or an amended CoC listed in 

§ 72.214. 

Paragraph 72.212(b)(5)(i) requires the 

general licensee to perform written 

evaluations which establish that the 

relevant cask, once loaded with spent 

fuel or once the changes authorized by 

an amended CoC have been applied, 

will conform to the terms, conditions, 

and specifications of a CoC or an 

amended CoC listed in § 72.214. 

Paragraph 72.212(b)(11) requires, among 

other things, that the general licensee 

comply with the terms, conditions, and 

specifications of the CoC or the 

amended CoC, as appropriate. Section 

72.214 lists the casks that are approved 

for storage of spent fuel under the 

conditions specified in their CoC. 

The licensee loaded spent fuel in the 

HI–STORM FW Storage System MPC– 

37, MPC 248, for storage in the ISFSI at 

STPEGS under CoC No. 1032, 

Amendment No. 2, under its general 

license. Condition 6 of the CoC states, 

‘‘Features or characteristics for the site 

or system must be in accordance with 

Appendix B to this certificate.’’ 

Appendix B, Section 3.3 of the CoC 

requires, with certain approved 

alternatives that are not relevant in this 

case, the HI–STORM FW MPC–37 to 

meet the American Society of 

Mechanical Engineers Boiler and 

Pressure Vessel Code, 2007 Edition 

(ASME Code). Section III, Subsection 

NB, of the ASME Code requires that 100 

percent of the weld seam joining the 

baseplate to the shell of the canister be 

examined by a radiography test (RT). 

Further, ASME Code Section III, 

Subsection NB requires, in part, that 

‘‘examination of a weld repair shall be 

repeated as required for the original 

weld.’’ Thus, in effect, the NRC staff is 

considering an exemption from the 

requirement to repeat volumetric 

examination by RT as required for the 

original weld on a 1-inch portion of the 

repaired weld. 

During a review of manufacturing 

documents, the manufacturer 

determined that a 1-inch section of the 

shell-to-baseplate weld on MPC 248 was 

not properly digitally radiographed after 

a weld repair. When notified of this 

issue, the licensee had already loaded 

MPC 248 with spent fuel assemblies and 

was in the process of preparing the MPC 

for long-term storage at the STPEGS 

ISFSI pad. The affected MPC is 

currently in a safe, analyzed condition 

in the STPEGS Unit 1 Fuel Handling 

Building cask decontamination area. 

This exemption would, if granted, 

permit the licensee to continue using 

MPC 248 to store spent fuel for the 

service life of the canister, including 

transferring the MPC to a HI–STORM 

FW overpack, without volumetric 

examination data from radiographic 

testing for a 1-inch section of the 

repaired weld seam joining the 

baseplate to the canister shell. In order 

for this exemption to exempt the 

licensee from all relevant provisions, 

the licensee would also need an 

exemption from 10 CFR 72.214. As the 

licensee did not request an exemption 

from 10 CFR 72.214, as part of the NRC 

staff’s consideration of the requested 

exemption, the NRC staff will also 

consider granting an exemption from 10 

CFR 72.214 upon its own initiative, in 

accordance with 10 CFR 72.7. For 

brevity, whenever this analysis refers to 

the requested exemption it means both 

the exemption requested by the licensee 

and the exemption from 10 CFR 72.214. 

III. Discussion 

Pursuant to 10 CFR 72.7, the 

Commission may, upon application by 

any interested person or upon its own 

initiative, grant such exemptions from 

the requirements of the regulations of 10 

CFR part 72 as it determines are 

authorized by law and will not endanger 

life or property or the common defense 

and security and are otherwise in the 

public interest. 

Authorized by Law 

Section 72.7 allows the NRC to grant 

exemptions from the requirements of 10 

CFR part 72. The NRC staff has 

determined that issuance of this 

exemption is consistent with the Atomic 

Energy Act of 1954, as amended, and 

not otherwise inconsistent with NRC’s 

regulations or other applicable laws. 

Therefore, the requested exemption is 

authorized by law. 

Will Not Endanger Life or Property or 

the Common Defense and Security 

This exemption would, if granted, 

exempt the licensee from the 

requirement to repeat volumetric 

examination as required for the original 

weld on a 1-inch portion of the repaired 

weld in ASME Code Section III, 

Subsection NB, which the licensee is 


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required to follow by the relevant 

technical specifications. If would also, if 

granted, exempt the licensee from the 10 

CFR 72.154(b) requirement to have 

available documentary evidence that 

material and equipment conform to the 

procurement specifications prior to 

installation or use of the material and 

equipment and to retain or have 

available this documentary evidence for 

the life of the ISFSI or spent fuel cask. 

The licensee supported this 

exemption request with a structural 

evaluation for the MPC and a separate 

structural analysis, both of which 

assumed a weld strength reduction 

factor of 0.8 to account for the missing 

RT examination. The structural 

evaluation showed that MPC 248 

maintains structural and confinement 

functions and that, even with the 0.8 

weld strength reduction factor, MPC 248 

would still meet the ASME Code, 

Section III, Subsection NB structural 

analysis requirements. The NRC’s 

review and evaluation of this 0.8 weld 

strength reduction factor and the 

licensee’s structural analysis for MPC 

248 are found in the Materials Review 

for the Requested Exemption and the 

Structural Review for the Requested 

Exemption section of this notice, 

respectively. 

Review of the Requested Exemption 

The HI–STORM FW storage system 

consists of a sealed metallic multipurpose canister (MPC) contained 

within an overpack constructed from a 

combination of steel and concrete. The 

HI–STORM FW overpack can be loaded 

with the MPC containing spent fuel 

using the HI–TRAC VW transfer cask 

and prepared for storage while inside 

the 10 CFR part 50 facility. The HI– 

TRAC VW transfer cask is required for 

shielding and protection of the spent 

fuel during loading and closure of the 

MPC and during movement of the 

loaded MPC from the cask loading area 

of a nuclear plant spent fuel pool to the 

storage overpack. The MPC enclosure 

vessels are cylindrical weldments with 

identical and fixed outside diameters. 

Each MPC is an assembly consisting of 

a honeycomb fuel basket, a baseplate, a 

canister shell, a lid, and a closure ring. 

The number of spent fuel storage 

locations in an MPC depends on the 

type of fuel assembly. The MPC–37 

model in use at STPEGS is designed to 

hold 37 pressurized water reactor fuel 

assemblies. 

The NRC has previously approved the 

HI–STORM FW storage system in CoC 

No. 1032, including Amendment No. 2 

to the CoC, which is the version of the 

CoC in use at STPEGS. The requested 

exemption does not change the 

fundamental design, components, 

contents, or safety features of the storage 

system. The NRC staff has evaluated the 

applicable potential safety impacts of 

granting the requested exemption to 

assess the requested exemption’s 

potential for danger to life or property 

or the common defense and security; the 

evaluation and resulting conclusions are 

presented in this notice. The potential 

impacts identified for this exemption 

request were in the areas of materials, 

structural integrity, and confinement 

capability. The staff did not identify any 

potential impacts in the areas of 

criticality, shielding, and thermal 

conditions. 

Materials Review for the Requested 

Exemption: The licensee asserted that 

although MPC 248 does not meet the 

ASME Code requirements specified in 

Appendix B, Section 3.3 of the CoC, 

MPC 248 continues to meet its safety 

functions. The licensee stated that after 

the completion of spent fuel loading, 

drying, and closure welding of MPC 

248, Holtec International, the CoC 

holder, informed the licensee that MPC 

248 does not fully meet the 

requirements in CoC Appendix B. More 

specifically, the Holtec HI–STORM FW 

MPC design and certification is based 

on compliance with ASME Code 

Section III, with certain approved 

alternatives. Portions of ASME Code 

Section III, Subsection NB–5000, require 

that weld repairs in the MPC 

confinement boundary be examined to 

the same criteria as the initial welds. 

Section III, Subsection NB also requires 

that 100 percent of the MPC shell-tobaseplate welds be volumetrically 

examined using RT, in accordance with 

Section III, NB–5230. 

During fabrication, Holtec performed 

a typical weld repair of the MPC 248 

shell-to-baseplate weld after the initial 

digital RT examination showed a 

section of the weld had lack of fusion. 

The weld was excavated to remove the 

lack-of-fusion defect and a successful 

liquid penetrant test (PT) examination 

of the entire excavated area was 

performed. The dimensions of the 

excavated area are approximately 9 

inches in length by 0.5 inches in width 

and 0.5 inches in depth (through wall at 

the defect location). The licensee stated 

that detailed profile dimensions of the 

repaired area are not available but 

referred to Holtec procedures that 

require a 3-to-1 taper for weld repair 

excavations. The weld repair was 

performed using an ASME Code Section 

IX qualified gas tungsten arc weld 

procedure and successfully passed a 

final PT exam. After the PT exam was 

completed, the unit was reinspected 

using the same digital RT process used 

on the original weld, but only 8 inches 

of the 9-inch length were examined. The 

missing 1-inch section is located at the 

end of one side of the excavated area. 

The licensee stated that the RT on the 

original weld did not identify weld 

defects on the end of the excavated area 

containing this 1-inch portion, but a 

portion of the weld in this 1-inch 

section had to be removed to access the 

defects in the adjacent portion of the 

shell-to-baseplate weld and allow repair 

welding to be performed. Following 

completion of the weld repair, MPC 248 

successfully passed a helium leakage 

test during factory acceptance testing as 

well as a hydrostatic test performed at 

STPEGS during loading operations. 

According to the licensee, the repairs 

along the MPC shell-to-base plate weld 

were completed per Holtec’s written 

procedures. After completing the 

repairs, Holtec examined the repaired 

area by PT and determined that the PT 

examination results met the acceptance 

criteria of ASME Code Section III, NB– 

5350. Holtec performed the post-repair 

RT examination and later determined 

that the RT examination which met the 

acceptance criteria of ASME Code 

Section III, NB–5320 included only 8 of 

the 9 inches. 

The licensee’s assertion that MPC 248 

continues to meet all its safety functions 

is based on the following: 

• The weld repair was performed in 

accordance with all Holtec quality 

procedures. 

• MPC 248 has 653 inches of welds 

in total. Holtec performed an RT 

examination on all those welds except 

for the approximately 1-inch section of 

repaired weld. This 1-inch section is 

approximately 0.15 percent of the MPC 

248 welds. The remaining 99.85 percent 

of MPC 248’s welds were fully 

inspected. 

To support its weld strength 

reduction factor, the licensee referenced 

the weld strength reduction factor of 0.8 

from NRC Interim Staff Guidance 

(ISG)—15, ‘‘Materials Evaluation,’’ for 

welded austenitic stainless steel spent 

fuel storage canisters that are examined 

using progressive, multiple-layer PT 

examinations in lieu of a volumetric 

examination nondestructive 

examination (NDE) method that is 

required by ASME Code Section III, 

Subsection NB. 

The licensee also reviewed the 

requirements in several sections of the 

ASME code to support its selection of 

the weld strength reduction factor value 

from ISG–15. Specifically, the licensee 

reviewed the joint efficiency values 

included in ASME Code Section VIII, 

Division 1 and Section III, Subsection 

ND. The licensee also reviewed the 


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quality factor for welded joints in ASME 

Code Section III, Subsection NG. The 

licensee also noted that the SA–240 

Type 304 stainless steel design stress 

values applicable to ASME Code 

Section VIII, Division 1 and Section III, 

Subsection ND are generally equal to the 

design stress intensity values applicable 

to Section III, Subsection NB—which 

apply to the weld in question—except 

for minor variances at 300 and 400 

degrees Fahrenheit. 

The licensee noted that ASME Code 

Section VIII, Division 1, which governs 

the design and construction of nonnuclear pressure vessels, specifies that 

Category C butt joints have a weld 

efficiency of 0.85 when subject to spot 

radiography, as specified in ASME 

Code, Section VIII, Division 1, UW–52. 

The licensee noted that spot 

radiography requires a minimum of one 

6-inch spot to be RT examined for every 

50-foot increment of the weld. The 

licensee stated that the Category C butt 

joints are more critical than Category C 

corner joints, which is the type of joint 

for the weld in question. The licensee 

also stated that by comparison, more 

than 99 percent of the MPC shell-tobaseplate weld was examined by RT, far 

exceeding the requirement for spot 

radiography per ASME Code Section 

VIII, Division 1, UW–52. Therefore, the 

licensee concluded that a 0.85 value for 

joint efficiency is conservative for 

evaluation of MPC 248, making the 

selected 0.8 value even more 

conservative. 

The licensee noted that ASME Code 

Section III, Subsection ND, which 

applies to Class 3 nuclear components, 

also specifies a joint efficiency of 0.85 

for Category C butt welds subject to spot 

radiography. The licensee stated that the 

inspection performed on the shell-tobase plate weld for MPC 248 exceeds 

these minimum ASME Code 

requirements for spot radiography in 

ASME Code Section III, Subsection ND 

because more than 99 percent of the 

weld was examined by RT. 

The licensee also compared the value 

of the weld strength reduction factor 

from ISG–15 to the requirements of 

ASME Code Section III, Subsection NG, 

which is applicable to core support 

structures of nuclear facility 

components. The licensee pointed to 

Table NG–3352–1, which specifies a 

quality factor for a welded joint of 0.75 

for a full penetration weld subjected to 

PT for both the root pass and the final 

pass. The licensee stated that the quality 

factor for welded joints in Table NG– 

3352–1 would be overly conservative 

because more than 99 percent of the 

shell-to-baseplate weld for MPC 248 was 

volumetrically examined using RT, 100 

percent of the weld received surface 

examination using PT, and the weld 

excavation cavity at 3-to-1 taper at the 

1-inch weld location received PT. This 

is discussed in detail in the staff’s 

independent analysis in this notice. 

The NRC staff reviewed the 

information provided by the licensee 

including: (1) The licensee’s 

comparisons of the weld strength 

reduction factor to the joint efficiency 

values based on requirements contained 

in ASME Code Section VIII, Division 1, 

and ASME Code Section III, Subsection 

ND; and to the quality factor for welded 

joints in ASME Code Section III, 

Subsection NG; (2) the specific 

requirements in those ASME Code 

sections; (3) the guidance in, and 

applicability of ISG–15; and (4) the 

information provided by the licensee 

regarding the weld repair procedures 

and post-weld repair NDE results. 

The staff determined that although the 

weld strength reduction factors 

specified in the ASME Code sections 

cited by the licensee are not applicable 

to the Holtec HI–STORM FW MPC–37— 

which was approved using the design 

and construction requirements in ASME 

Code, Section III, Subsection NB—the 

values are conservative with respect to 

a possible weld strength reduction 

factor for MPC 248 because more than 

99 percent of the shell-to-baseplate weld 

was examined using RT and 100 percent 

was examined using PT. As discussed in 

this notice, the staff calculated two 

potential weld strength reduction 

factors, both are which are conservative. 

Both calculated values are greater than 

the licensee’s 0.8 value, making the 

licensee’s value more conservative. In 

addition, the staff notes that: (1) Only 

the fraction of the 1-inch-long three-toone tapered section of the weld that was 

removed was not examined by RT after 

the repair; (2) the portion of the 1-inch 

weld that remained after excavation at 

the three-to-one tapered section was 

volumetrically examined by RT prior to 

excavation and met the acceptance 

criteria of ASME Code Section III, NB– 

5320; (3) 100 percent of the repair weld 

section was successfully examined by 

PT both after excavation and after 

repair; and (4) more than 88 percent of 

the approximately 9-inch repair weld 

section was examined using RT. 

Therefore, the staff concluded that the 

values of the weld strength reduction 

factors derived from the ASME Code 

sections cited by the licensee 

conservatively bound the reduction in 

the weld strength of the shell-tobaseplate weld of MPC 248 as a result 

of possible weld defects in the 1-inch 

portion of the repair weld that was not 

examined by RT. 

The staff also reviewed the guidance 

in ISG–15 which states that, if 

progressive surface examinations (i.e., 

sequential examinations conducted as a 

multi-pass weld is deposited) such as 

multiple layer PT or magnetic particle 

testing are used for a spent fuel storage 

canister closure lid weld in lieu of a 

volumetric examination, a weld strength 

reduction factor of 0.8 is to be imposed 

on the weld design to account for 

imperfections or flaws that may have 

been missed by the progressive surface 

examinations. The staff determined that, 

although the guidance for the use of the 

weld strength reduction factor in ISG– 

15 was not intended to be applied for 

an MPC shell-to-baseplate weld, the 

value of the weld strength reduction 

factor from ISG–15 would be 

conservative for the MPC 248 shell-tobaseplate weld for the same reasons 

provided for the comparisons of the 

weld strength reduction factors from the 

ASME Code sections cited by the 

licensee and discussed in the previous 

paragraph. Therefore, the staff 

concluded that the values of the weld 

strength reduction factor from ISG–15 

conservatively bound the reduction in 

the weld strength of the shell-tobaseplate weld of MPC 248 as a result 

of possible buried weld defects in the 1- 

inch portion of the repair weld that was 

not examined by RT. 

The NRC staff conducted an 

independent analysis of MPC 248 

considering the MPC materials and the 

design of the shell-to-baseplate weld. 

The staff’s analysis postulated that the 

portion of the repaired area of the weld 

that was not subjected to the post-weld 

repair RT examination includes a buried 

weld flaw. 

The NRC staff used this initial 

postulate because: (1) The portion of the 

original weld in the 1-inch section was 

examined by PT after the weld 

excavation’s completion; and (2) the 

completed repair weld was also PTexamined. Both of these examinations 

reveal no surface-breaking flaws, 

indicating that if a flaw was to exist in 

that 1-inch section, it would be a buried 

weld flaw. The staff determined that for 

the entire shell-to-baseplate weld, the 

weld strength reduction factor that 

would be applied to the structural 

analysis of such a joint to account for a 

buried weld flaw per the ASME Code 

would be at least 0.99 because: (1) The 

entire section of the shell-to-base plate 

weld and the section of the repair weld 

that was RT-examined were verified to 

be free of any relevant flaws; (2) the 

design of the MPC shell and MPC 

baseplate are sufficiently thick and 

provide sufficient stiffness to the MPC 

shell to prevent significant stress 


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concentrations for relatively small 

buried weld flaws; (3) the MPC shell, 

baseplate, and the shell-to-baseplate 

weld are all high toughness materials 

that are not susceptible to brittle 

fracture; and (4) MPC 248 successfully 

passed a helium leakage test during 

factory acceptance testing and a 

hydrostatic pressure test during the 

loading operations. This number does 

not credit the 1-inch section without 

RT. The NRC staff calculated this 

number dividing the length of the 

section that that did not receive RT by 

the total length of the shell-to-baseplate 

weld and then subtracting that result 

from 1. This method produces a weld 

strength reduction factor that is greater 

than .99. Given that the licensee’s 

selected weld strength reduction factor 

of 0.8 is less than this staff-calculated 

value, the licensee’s factor accounts for 

a greater reduction in weld strength due 

to a buried flaw. 

In addition to the above analysis, the 

staff conducted a weld strength 

reduction factor analysis using greater 

conservatisms. Specifically, the staff 

assumed a worst-case flaw size that 

considered information provided by the 

licensee on the results of the initial RT 

of the shell-to-baseplate weld; the 

profile of the weld excavation and the 

weld repair process; and the NDE 

conducted after excavation and again 

after the weld repair was completed. 

This calculation based the weld strength 

reduction factor on only the repaired 

weld rather than the entire shell-tobaseplate weld. In this evaluation, the 

staff also did not credit the presence of 

the entire 1-inch repair weld which was 

not RT-examined post repair. This 

calculation would be conservative 

relative to the actual reduction in weld 

strength because the 1-inch portion of 

the weld that did not receive post-repair 

RT was initially examined by RT per 

ASME Code Section III, NB–5230 and 

shown to meet the ASME Code Section 

III, NB–5320 acceptance criteria prior to 

weld excavation (as previously 

discussed, this section of the weld is 

located within the 3-to-1 taper area). 

Additionally, the weld excavation 

cavity and post-repair weld were both 

PT examined per ASME Code Section 

III, NB–5230 and met the acceptance 

criteria of ASME Code Section III, NB– 

5350. In this case, the 1-inch section is 

11 percent of the 9-inch repair section. 

Thus, the same calculation discussed 

above produces a weld strength 

reduction factor of 0.89. 

Given that the licensee’s selected 

weld strength reduction factor of 0.8 is 

less than both of the staff-calculated 

values, it would account for a greater 

reduction in weld strength due to a 

buried flaw than either of those values. 

Therefore, the 0.80 weld strength 

reduction factor is conservative. 

The staff’s independent analyses of 

the weld strength reduction factor for 

MPC 248 are conservative because: (1) 

The weld repair procedure with the 

multi-pass manual gas tungsten arc 

weld was developed to facilitate a weld 

repair, provide more control over weld 

deposition, and minimize the 

introduction of weld flaws; (2) the 1- 

inch weld is within the three-to-one 

taper section of the repair excavation 

with sound weld metal backing based 

on the initial RT results and the weld 

excavation cavity PT results prior to the 

weld repair; (3) the post-repair weld 

examinations using PT and RT met the 

acceptance criteria in ASME Code 

Section III, NB–5300; (4) any weld 

repair flaw present in the non-examined 

RT weld repair section would be limited 

to the dimensions of the weld repair in 

the tapered area of the excavation; (5) 

based on the post-repair PT results, any 

flaw introduced during repair welding 

would be embedded in the weld with 

low stress concentration of little to no 

significance to structural performance or 

the confinement function of the MPC; 

and (6) the staff’s analysis was based on 

a maximum of 1-inch missing weld in 

the MPC shell-to-baseplate weld. 

Based on the points above, any weld 

flaw present in the 1-inch section that 

was not examined by RT after the weld 

repair would be a small relative to the 

length of that section of the weld. The 

staff’s analysis is conservative because, 

as stated above, the analysis assumed no 

credit for the entire portion of the weld 

that was not examined by RT after the 

repair. Because the licensee’s 0.8 weld 

strength reduction factor is more 

conservative than the values of the weld 

strength reduction factor the staff 

calculated, the staff’s independent 

analysis shows that the weld strength 

reduction factor of 0.8 used by the 

licensee is sufficient to account for the 

possible presence of non-surface 

breaking flaws in the portion of the 

repair weld that was not subjected to 

post-repair volumetric examination. 

Therefore, the staff finds the 0.8 weld 

strength reduction factor acceptable. 

The licensee’s structural analysis using 

this weld strength reduction factor is 

analyzed in this notice. 

Evaluation Findings of Materials 

Review: As a result of the analyses 

discussed above, the NRC staff finds 

that the weld strength reduction factor 

provided by the licensee is sufficient to 

account for the presence of undetected 

flaws that may be present in the shellto-baseplate weld of MPC 248, loaded 

under CoC No. 1032, Amendment No. 2. 

Therefore, the use of a 0.8 weld strength 

reduction factor in the structural 

evaluation would not endanger life or 

property or the common defense and 

security if the requested exemption 

were granted. 

Structural Review for the Requested 

Exemption: The staff’s structural review 

focused on the re-analysis of the shellto-baseplate weld, as provided in 

Enclosure 2 (proprietary), ‘‘HI–STORM 

FW MPC Stress Analysis,’’ of the 

exemption request, to verify that the 

safety function of the MPC is 

maintained after considering a weld 

strength reduction factor to the 

allowable stress values used as design 

criteria. As discussed above, the 

licensee applied a weld strength 

reduction factor in its analysis to 

account for imperfections or flaws that 

may be missed for the 1-inch weld 

portion without post-repair RT. 

Re-Analysis of the Shell-to-Baseplate 

Weld: The HI–STORM FW Final Safety 

Analysis Report (FSAR), HOLTEC 

Report No. HI–2114830, Table 10.1.4, 

‘‘HI–STORM FW MPC NDE 

Requirements,’’ establishes the weld 

acceptance criteria that provide 

reasonable assurance that the weld will 

perform its design function under all 

loading conditions as defined in ASME 

Code, Section III, Subsection NB. In 

accordance with Appendix B, Section 

3.3, ‘‘Codes and Standards,’’ of CoC No. 

1032, the HI–STORM FW MPC–37 must 

meet the 2007 Edition of the ASME 

Code. The ASME Code Section III, 

Subsection NB, states, in part, that 

‘‘examination of a weld repair shall be 

repeated as required for the original 

weld.’’ For original welds, it is required 

that 100 percent of the weld seam 

joining the baseplate to the shell of the 

canister be examined by RT. Since the 

unexamined portion of the repair weld 

is not in conformance with the ASME 

Code requirements described in the 

CoC, the licensee’s structural evaluation 

seeks to demonstrate that the use of the 

affected MPC 248 will not adversely 

impact its structural safety function 

after considering a weld strength 

reduction factor used to account for the 

non-conformance condition. 

As discussed above in the materials 

review of the requested exemption, the 

staff concluded that the licensee’s weld 

strength reduction factor of 0.8 (i.e., an 

overall 20 percent reduction in the 

allowable stress) is sufficient to account 

for potential imperfections or flaws that 

may have been missed by an incomplete 

RT when considering the size of the 

unexamined portion of the repair weld. 

The licensee applied this weld strength 

reduction factor to the allowable stress 

intensity used in the five load cases 


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Federal Register / Vol. 87, No. 83 / Friday, April 29, 2022 / Notices 25559 

identified as the governing load 

combinations for the MPC 248 shell-tobaseplate weld per the HI–STORM FW 

FSAR (HOLTEC Report No. HI– 

2114830, Revision 5) to re-evaluate the 

safety factors that are available and 

demonstrate that the design function 

will be maintained. The five load cases 

are as follows: The design condition 

with a 120 pounds per square inch 

gauge (psig) normal internal pressure 

only to bound short-term normal 

operations (Case 1), an accident 

condition with a 200 psig accident 

internal pressure (Case 2), a short-term 

MPC lifting operation with a 120 psig 

operating internal pressure plus weight 

of the contents (Case 3), an off-normal 

condition with a 120 psig off-normal 

internal pressure plus bounding offnormal temperature contours (Case 4), 

and a design basis short-term operation 

with a 120 psig internal pressure plus 

bounding short-term operation 

temperature contours (Case 5). By 

comparing the reduced allowable stress 

of each loading condition to the 

resultant stress obtained from the finite 

element analysis performed by the 

licensee in the structural analysis of the 

HI–STORM FW system (Holtec Report 

HI–2094418, Revision 20), the licensee 

calculated a new safety factor for each 

loading condition. The analysis 

demonstrated that the shell-to-baseplate 

weld maintains a safety factor above 1.0 

for all loading conditions and that 

sufficient design margin remains to 

accommodate the resultant stress from 

each loading condition even with the 

reduced stress allowable used to 

account for potential imperfections or 

flaws in the repaired weld. The licensee 

further stated that, in addition to the 

weld strength reduction factor, the 

analysis also retains several 

conservatisms from the existing FSAR 

design basis analysis, such as using 

bounding pressures, temperatures, and 

temperature contours. 

While the NRC staff is not basing its 

conclusions on these conservatisms, the 

NRC staff notes that the use of these 

conservative values in the analysis 

demonstrate that additional design 

margin remains available to 

accommodate resultant stress. 

Evaluation Findings of Structural 

Review: The NRC staff reviewed the 

analysis performed in Enclosure 2 

(proprietary) of the exemption request 

for the MPC shell-to-baseplate weld and 

finds that the licensee evaluation 

demonstrates that a safety factor greater 

than 1.0 is maintained (i.e., calculated 

stresses remain below the allowable 

stress intensities with the reduction 

factor) for all normal, off-normal, and 

accident conditions after the stress 

allowable for each load case is reduced 

by 20 percent to account for 

imperfections or flaws that may be 

missed due to the non-conforming weld 

inspection. The staff notes that the use 

of a weld strength reduction factor to 

the allowable stress values is similar to 

other approved alternatives to the 

ASME code examination requirements 

as described in NUREG–2215, 

‘‘Standard Review Plan for Spent Fuel 

Dry Storage Systems and Facilities,’’ to 

account for imperfections or flaws that 

may be missed by other examinations. 

While the alternatives described in 

NUREG–2215 are not applicable to this 

weld, as discussed above in the 

materials section, the NRC finds their 

use acceptable in this instance. During 

its review, the staff also verified that the 

licensee has properly applied the weld 

strength reduction factor of 0.8 to 

applicable allowable stress values for 

the design criteria. The staff also notes 

that no potential for stress cycling is 

expected at the unexamined portion of 

the repair. As discussed in Section 

3.1.2.5 of CoC No. 1032 FSAR, fatigue 

failure is not a credible concern for the 

MPC since it is not an active system 

(i.e., no moving parts) and is not subject 

to significant stress cycling due to rapid 

temperature changes or significant 

pressure changes. Therefore, there is no 

credible concern of fatigue failure if any 

flaw introduced during the weld repair 

is considered. 

As set forth above, the licensee has 

demonstrated that the shell-to-baseplate 

weld for MPC 248, loaded under CoC 

No. 1032, Amendment No. 2, is capable 

of maintaining its structural integrity 

and performing its safety function under 

normal, off-normal, and accident 

conditions. Therefore, the staff 

concludes that the structural properties 

of MPC 248, as addressed in the 

exemption request, remain in 

compliance with 10 CFR part 72, and 

therefore, from a structural perspective, 

this exemption, if granted, would not 

endanger life or property or the common 

defense and security. 

Confinement Review for the 

Requested Exemption: The licensee 

stated on page 1 of Enclosure 1 of its 

exemption request that MPC 248 

successfully passed a helium leakage 

test during factory acceptance testing 

following completion of the weld repair, 

as well as a hydrostatic test which was 

performed at STPEGS during loading 

operations. According to the licensee, 

the helium leakage test performed on 

MPC 248 was in conformance with the 

FSAR and the applicable Technical 

Specifications for the HI–STORM FW 

storage system and satisfied the 

‘‘leaktight’’ criteria in ANSI N14.5– 

1997. 

Evaluation Findings of Confinement 

Review: The staff found that, because 

MPC 248 successfully passed a helium 

leakage test during a fabrication 

acceptance test following completion of 

the weld repair, the MPC meets the 

leaktight criteria of ANSI N14.5–1997. 

Further, MPC 248 passed a hydrostatic 

test performed at STPEGS during 

loading operations, which provides 

further evidence of no discernable 

leakage from this MPC at the time of 

loading. The staff therefore concludes 

that MPC 248 meets the regulatory 

requirements for confinement in 10 CFR 

part 72 and, therefore, the weld repair 

completed on MPC 248 has had no 

effect on the confinement performance 

of the MPC in question. Consequently, 

from a confinement perspective, this 

exemption, if granted, would not 

endanger life or property or the common 

defense and security. 

Conclusion Regarding Deviation from 

Weld Inspection Requirement: As noted 

above, the NRC staff did not identify 

any potential effects on criticality, 

shielding, and thermal conditions. 

Therefore, based on that fact and the 

above discussions, the NRC staff 

concludes that an exemption exempting 

the licensee from the requirement to 

repeat volumetric examination for the 1- 

inch portion of the repaired weld, if 

granted, would not endanger life or 

property or the common defense and 

security. 

Record Keeping Provision Evaluation: 

As noted above, the licensee also 

requested an exemption from the 10 

CFR 72.154(b) requirement to have 

available documentary evidence that 

material and equipment conform to the 

procurement specifications prior to 

installation or use of the material and 

equipment and to retain or have 

available this documentary evidence for 

the life of the ISFSI or spent fuel cask. 

The records covered by the requested 

exemption are the records detailing the 

results for the RT discussed above. As 

previously detailed, the NRC staff has 

concluded that exempting the licensee 

from the requirement to repeat 

volumetric examination as required for 

the original weld on a 1-inch portion of 

the repaired weld would not endanger 

life or property or the common defense 

and security. If not performing the RT 

does not endanger life or property or the 

common defense and security, it follows 

that not retaining records of those test 

results would also not endanger life or 

property or the common defense and 

security. Therefore, the NRC staff finds 

that the requested exemption from 10 

CFR 72.154(b), if granted, would not 


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endanger life or property or the common 

defense and security. 

Otherwise in the Public Interest 

In considering whether granting the 

requested exemption is in the public 

interest, the NRC staff considered the 

alternative of not granting the requested 

exemption. If the requested exemption 

were not granted, in order to comply 

with the CoC, MPC 248 would need to 

be opened and unloaded, the contents 

loaded in new MPC, and the new MPC 

welded and tested. This option would 

entail a higher risk of canister handling 

accidents, additional personnel 

exposure, and greater cost to the 

licensee. This option would also 

generate additional radioactive 

contaminated material and waste from 

operations. For example, the lid would 

have to be removed, which would 

generate cuttings from removing the 

weld material that could require 

disposal as contaminated material. This 

radioactive waste would be transported 

and ultimately disposed of at a qualified 

low-level radioactive waste disposal 

facility, potentially exposing it to the 

environment. 

Further, data subject to the requested 

exemption from 10 CFR 72.154(b) is the 

data that comes from the test from 

which the licensee is being exempted. 

Without the data from the test, the 

licensee cannot satisfy 10 CFR 

72.154(b). Thus, granting an exemption 

from the test requirements but not from 

the record-keeping requirement would 

still force the license to open and 

unload MPC 248, load the contents in 

new MPC, and weld and test the new 

MPC, meaning all the potential negative 

effects would still occur. 

Based on the above, approving the 

requested exemption reduces the 

opportunity for a release of radioactive 

material compared to the alternative to 

the proposed action because there will 

be no operations involving the opening 

of the MPC that confines the spent 

nuclear fuel, potentially exposing 

radioactive waste to the environment. It 

will also generate less radioactive waste 

for disposal. Thus, the proposed 

exemption is consistent with NRC’s 

mission to protect public health and 

safety. Therefore, the requested 

exemption is otherwise in the public 

interest. 

Environmental Consideration 

The NRC staff also considered in the 

review of this exemption request 

whether there would be any significant 

environmental impacts associated with 

the exemption. The NRC staff 

determined that this proposed action 

fits a category of actions that do not 

require an environmental assessment or 

environmental impact statement. 

Specifically, the requested exemption 

meets the categorical exclusion in 10 

CFR 51.22(c)(25). 

Granting an exemption from 10 CFR 

72.212(a)(2), 10 CFR 72.212(b)(3), 10 

CFR 72.212(b)(5)(i), 10 CFR 

72.212(b)(11), and 10 CFR 72.214 would 

only relieve the licensee from the 

inspection requirement found in TS 3.3 

of Attachment B of CoC No. 1032. With 

this requested exemption, the licensee 

would be exempt from the requirement 

to repeat volumetric examination as 

required for the original weld on a 1- 

inch portion of the repaired weld 

joining the canister baseplate to the 

canister shell of the HI–STORM FW 

MPC 248. Granting an exemption from 

10 CFR 72.154(b) only relieves the 

licensee from the recordkeeping 

requirement associated with retaining 

and having available documentary 

evidence of a complete volumetric 

examination of the subject weld. A 

categorical exclusion for inspection 

requirements is provided under 10 CFR 

51.22(c)(25)(vi)(C), and a categorical 

exclusion for recordkeeping 

requirements is provided under 10 CFR 

51.22(c)(25)(A). In both cases, the 

criteria in 10 CFR 51.22(c)(25)(i)–(v) 

must also be satisfied. 

In its review of the exemption request, 

the NRC staff determined, that, in 

accordance with 10 CFR 51.22(c)(25): (i) 

Granting the exemption does not 

involve a significant hazards 

considerations because granting the 

exemption neither reduces a margin of 

safety, creates a new or different kind of 

accident from any accident previously 

evaluated, nor significantly increases 

either the probability or consequences 

of an accident previously evaluated; (ii) 

granting the exemption would not 

produce a significant change in either 

the types or amounts of any effluents 

that may be released offsite because the 

requested exemption neither changes 

the effluents nor produces additional 

avenues of effluent release; (iii) granting 

the exemption would not result in a 

significant increase in either 

occupational radiation exposure or 

public radiation exposure, because the 

requested exemption neither introduces 

new radiological hazards nor increases 

existing radiological hazards; (iv) 

granting the exemption would not result 

in a significant construction impact, 

because there are no construction 

activities associated with the requested 

exemption; and (v) granting the 

exemption would not increase either the 

potential for or consequences from 

radiological accidents because, even 

with the exemption, the canister will 

still be bounded by the FSAR analysis 

and will remain leaktight, and the 

exemption creates no new accident 

precursors at the STP ISFSI. Finally, as 

previously noted this exemption request 

involves recordkeeping requirements 

and inspection requirements under 10 

CFR 51.22(c)(25)(A) and (C), 

respectively. Accordingly, the requested 

exemption meets the criteria for a 

categorical exclusion in 10 CFR 

51.22(c)(25)(vi)(C). 

IV. Availability of Documents 

The documents identified in the 

following table are available to 

interested persons through one or more 

of the previously described methods. 

Document ADAMS 

accession No. 

South Texas Project, Units 1 and 2, Docket Nos. 50–498; 50–499; 72–1041, Independent Spent Fuel Storage Installation, 

Request for Exemption from Certificate of Compliance, Inspection Requirement for One Multipurpose Canister, dated 

March 11, 2022.

ML22070B140 

Request for Additional Information for Review of the South Texas Project Electric Generating Station Independent Spent 

Fuel Storage Installation, License No. SNM–2514, dated March 31, 2022.

ML22089A085 

South Texas Project, Units 1 and 2, Docket Nos. 50–498; 50–499; 72–1041, Independent Spent Fuel Storage Installation, 

Supplement to Request for Exemption from Certificate of Compliance (CoC) Inspection Requirement for One Multipurpose Canister, dated April 1, 2022.

ML22091A308 

Spent Fuel Project Office Interim Staff Guidance–15, Materials Evaluation, Revision 0, January 10, 2001 ............................. ML010100170 

Issuance of Certificate of Compliance No. 1032, Amendment No. 2 for the HI–STORM Flood/Wind Multipurpose Canister 

Storage System.

ML16280A008 * 

CoC No. 1032, Amendment No. 2 [Letter to K. Manzione re: Issuance of Certificate of Compliance No. 1032, Amendment 

No. 2 for the HI–STORM Flood/Wind Multipurpose Canister Storage System].

ML16280A017 


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Federal Register / Vol. 87, No. 83 / Friday, April 29, 2022 / Notices 25561 

Document ADAMS 

accession No. 

Certificate of Compliance No. 1032, Appendix B [Letter to K. Manzione re: Issuance of Certificate of Compliance No. 1032, 

Amendment No. 2 for the HI–STORM Flood/Wind Multipurpose Canister Storage System].

ML16280A019 

HI–2114830, Rev. 5, ‘‘Final Safety Analysis Report on the HI–STORM FW FSAR MPC Storage System’’ ............................. ML17179A444 

NUREG–2215, ‘‘Standard Review Plan for Spent Fuel Dry Storage Systems and Facilities’’ ................................................... ML20121A190 

* (Package). 

V. Conclusion 

Based on the foregoing 

considerations, the NRC staff has 

determined that, pursuant to 10 CFR 

72.7, the exemption is authorized by 

law, will not endanger life or property 

or the common defense and security, 

and is otherwise in the public interest. 

Therefore, the NRC grants the licensee 

an exemption from the requirements of 

10 CFR 72.212(a)(2), 10 CFR 

72.212(b)(3), 10 CFR 72.212(b)(5)(i), 10 

CFR 72.212(b)(11), and 10 CFR 72.214 

only with regard to meeting the 

requirement to repeat volumetric 

examination as required for the original 

weld on a 1-inch portion of the repaired 

weld in conformance with Section III, 

Subsection NB, of the ASME Code, 2007 

Edition, and 10 CFR 72.154(b) only with 

regard to maintaining and having 

available documentary evidence of the 

test for the service life of the canister. 

This exemption is effective upon 

issuance. 

Dated: April 25, 2022. 

For the Nuclear Regulatory Commission. 

Yoira K. Diaz-Sanabria, 

Chief, Storage and Transportation Licensing 

Branch, Division of Fuel Management, Office 

of Nuclear Material Safety and Safeguards. 

[FR Doc. 2022–09171 Filed 4–28–22; 8:45 am] 

BILLING CODE 7590–01–P 

NUCLEAR REGULATORY 

COMMISSION 

[NRC–2022–0001] 

Sunshine Act Meetings 

TIME AND DATE: Weeks of May 2, 9, 16, 

23, 30, June 6, 2022. The schedule for 

Commission meetings is subject to 

change on short notice. The NRC 

Commission Meeting Schedule can be 

found on the internet at: https://

www.nrc.gov/public-involve/publicmeetings/schedule.html. 

PLACE: The NRC provides reasonable 

accommodation to individuals with 

disabilities where appropriate. If you 

need a reasonable accommodation to 

participate in these public meetings or 

need this meeting notice or the 

transcript or other information from the 

public meetings in another format (e.g., 

Braille, large print), please notify Anne 

Silk, NRC Disability Program Specialist, 

at 301–287–0745, by videophone at 

240–428–3217, or by email at 

Anne.Silk@nrc.gov. Determinations on 

requests for reasonable accommodation 

will be made on a case-by-case basis. 

STATUS: Public. 

Members of the public may request to 

receive the information in these notices 

electronically. If you would like to be 

added to the distribution, please contact 

the Nuclear Regulatory Commission, 

Office of the Secretary, Washington, DC 

20555, at 301–415–1969, or by email at 

Wendy.Moore@nrc.gov or 

Betty.Thweatt@nrc.gov. 

MATTERS TO BE CONSIDERED: 

Week of May 2, 2022 

There are no meetings scheduled for 

the week of May 2, 2022. 

Week of May 9, 2022—Tentative 

Tuesday, May 10, 2022 

9:00 a.m. Strategic Programmatic 

Overview of the Fuel Facilities and 

the Spent Fuel Storage and 

Transportation Business Lines 

(Public) (Contact: Kellee Jamerson: 

301–415–7408) 

Additional Information: The meeting 

will be held in the Commissioners’ 

Conference Room, 11555 Rockville Pike, 

Rockville, Maryland. The public is 

invited to attend the Commission’s 

meeting live by webcast at the Web 

address—https://video.nrc.gov/. 

Thursday, May 12, 2022 

10:00 a.m. Briefing on Advanced 

Reactors Activities with Federal 

Partners (Public) (Contact: Caty 

Nolan: 301–415–1535) 

Additional Information: The meeting 

will be held in the Commissioners’ 

Conference Room, 11555 Rockville Pike, 

Rockville, Maryland. The public is 

invited to attend the Commission’s 

meeting live by webcast at the Web 

address—https://video.nrc.gov/. 

Week of May 16, 2022—Tentative 

There are no meetings scheduled for 

the week of May 16, 2022. 

Week of May 23, 2022—Tentative 

There are no meetings scheduled for 

the week of May 23, 2022. 

Week of May 30, 2022—Tentative 

Wednesday, June 1, 2022 

10:00 a.m. Transformation at the 

NRC—Sustaining Progress as 

Modern, Risk-Informed Regulator 

(Public) (Contact: Aida RiveraVarona: 301–415–4001) 

Additional Information: The meeting 

will be held in the Commissioners’ 

Conference Room, 11555 Rockville Pike, 

Rockville, Maryland. The public is 

invited to attend the Commission’s 

meeting live by webcast at the Web 

address—https://video.nrc.gov/. 

Friday, June 3, 2022 

10:00 a.m. Meeting with Advisory 

Committee on Reactor Safeguards 

(Public) (Contact: Larry Burkhart: 

301–287–3775) 

Additional Information: The meeting 

will be held in the Commissioners’ 

Conference Room, 11555 Rockville Pike, 

Rockville, Maryland. The public is 

invited to attend the Commission’s 

meeting live by webcast at the Web 

address—https://video.nrc.gov/. 

Week of June 6, 2022—Tentative 

There are no meetings scheduled for 

the week of June 6, 2022. 

CONTACT PERSON FOR MORE INFORMATION: 

For more information or to verify the 

status of meetings, contact Wesley Held 

at 301–287–3591 or via email at 

Wesley.Held@nrc.gov. 

The NRC is holding the meetings 

under the authority of the Government 

in the Sunshine Act, 5 U.S.C. 552b. 

Dated: April 27, 2022. 

For the Nuclear Regulatory Commission. 

Wesley W. Held, 

Policy Coordinator, Office of the Secretary. 

[FR Doc. 2022–09356 Filed 4–27–22; 4:15 pm] 

BILLING CODE 7590–01–P 

OFFICE OF PERSONNEL 

MANAGEMENT 

Comment Request for Review of a 

Generic Information Collection: 

Program Services Evaluation Surveys 

AGENCY: Office of Personnel 

Management. 


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