[Federal Register Volume 87, Number 83 (Friday, April 29, 2022)]
[Notices]
[Pages 25554-25561]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2022-09171]


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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.

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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:
    <bullet> 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: <a href="/cdn-cgi/l/email-protection#85d6f1e4e6fcabd6e6edf0e8e4ebebc5ebf7e6abe2eaf3"><span class="__cf_email__" data-cfemail="6d3e190c0e14433e0e0518000c03032d031f0e430a021b">[email protected]</span></a>. For technical 
questions, contact the individual listed in the For Further Information 
Contact section of this document.
    <bullet> 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 <a href="/cdn-cgi/l/email-protection#78283c2a562a1d0b170d0a1b1d38160a1b561f170e"><span class="__cf_email__" data-cfemail="cd9d899fe39fa8bea2b8bfaea88da3bfaee3aaa2bb">[email protected]</span></a>. For the convenience of the reader, 
instructions about obtaining materials referenced in this document are 
provided in the ``Availability of Documents'' section.
    <bullet> 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 
<a href="/cdn-cgi/l/email-protection#510115037f0334223e24233234113f23327f363e27"><span class="__cf_email__" data-cfemail="c9998d9be79bacbaa6bcbbaaac89a7bbaae7aea6bf">[email protected]</span></a> or call 1-800-397-4209 or 301-415-4737, between 
8:00 a.m. and 4:00 p.m.

[[Page 25555]]

(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: 
<a href="/cdn-cgi/l/email-protection#d397bcbdb2bfb7fd9bb2b1bab193bda1b0fdb4bca5"><span class="__cf_email__" data-cfemail="fbbf94959a979fd5b39a999299bb958998d59c948d">[email protected]</span></a>.

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 Sec.  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 Sec.  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

[[Page 25556]]

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 multi-
purpose 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-to-baseplate 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:
    <bullet> The weld repair was performed in accordance with all 
Holtec quality procedures.
    <bullet> 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

[[Page 25557]]

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 non-nuclear 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-to-baseplate 
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-to-base 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-to-one 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-to-baseplate 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-to-baseplate 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-to-baseplate 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 PT-examined. 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

[[Page 25558]]

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-to-baseplate 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 shell-to-
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 shell-to-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

[[Page 25559]]

identified as the governing load combinations for the MPC 248 shell-to-
baseplate 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 off-normal 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

[[Page 25560]]

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,      ML22070B140
 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.
Request for Additional Information for   ML22089A085
 Review of the South Texas Project
 Electric Generating Station
 Independent Spent Fuel Storage
 Installation, License No. SNM-2514,
 dated March 31, 2022.
South Texas Project, Units 1 and 2,      ML22091A308
 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.
Spent Fuel Project Office Interim Staff  ML010100170
 Guidance-15, Materials Evaluation,
 Revision 0, January 10, 2001.
Issuance of Certificate of Compliance    ML16280A008 *
 No. 1032, Amendment No. 2 for the HI-
 STORM Flood/Wind Multipurpose Canister
 Storage System.
CoC No. 1032, Amendment No. 2 [Letter    ML16280A017
 to K. Manzione re: Issuance of
 Certificate of Compliance No. 1032,
 Amendment No. 2 for the HI-STORM Flood/
 Wind Multipurpose Canister Storage
 System].

[[Page 25561]]

 
Certificate of Compliance No. 1032,      ML16280A019
 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].
HI-2114830, Rev. 5, ``Final Safety       ML17179A444
 Analysis Report on the HI-STORM FW
 FSAR MPC Storage System''.
NUREG-2215, ``Standard Review Plan for   ML20121A190
 Spent Fuel Dry Storage Systems and
 Facilities''.
------------------------------------------------------------------------
* (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


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