83_FR_47373 83 FR 47192 - Xcel Energy, Monticello Nuclear Generating Plant; Independent Spent Fuel Storage Installation

83 FR 47192 - Xcel Energy, Monticello Nuclear Generating Plant; Independent Spent Fuel Storage Installation

NUCLEAR REGULATORY COMMISSION

Federal Register Volume 83, Issue 181 (September 18, 2018)

Page Range47192-47203
FR Document2018-20283

The U.S. Nuclear Regulatory Commission (NRC) is issuing an exemption in response to a request submitted by Xcel Energy on October 18, 2017, from meeting Technical Specification (TS) 1.2.5 of Attachment A of Certificate of Compliance (CoC) No. 1004, Amendment No. 10, which requires that all dry shielded canister (DSC) closure welds, except those subjected to full volumetric inspection, be dye penetrant tested in accordance with the requirements of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code Section III, Division 1, Article NB-5000. This exemption applies to five loaded Standardized NUHOMS[supreg] 61BTH, Dry Shielded Canisters (DSCs) 11 through 15, at the Monticello Nuclear Generating Plant (MNGP) Independent Spent Fuel Storage Installation (ISFSI).

Federal Register, Volume 83 Issue 181 (Tuesday, September 18, 2018)
[Federal Register Volume 83, Number 181 (Tuesday, September 18, 2018)]
[Notices]
[Pages 47192-47203]
From the Federal Register Online  [www.thefederalregister.org]
[FR Doc No: 2018-20283]


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NUCLEAR REGULATORY COMMISSION

[Docket Nos. 72-58 and 50-263; NRC-2018-0207]


Xcel Energy, Monticello Nuclear Generating Plant; 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 Xcel Energy on October 
18, 2017, from meeting Technical Specification (TS) 1.2.5 of Attachment 
A of Certificate of Compliance (CoC) No. 1004, Amendment No. 10, which 
requires that all dry shielded canister (DSC) closure welds, except 
those subjected to full volumetric inspection, be dye penetrant tested 
in accordance with the requirements of American Society of Mechanical 
Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code Section III, 
Division 1, Article NB-5000. This exemption applies to five loaded 
Standardized NUHOMS[supreg] 61BTH, Dry Shielded Canisters (DSCs) 11 
through 15, at the Monticello Nuclear Generating Plant (MNGP) 
Independent Spent Fuel Storage Installation (ISFSI).

ADDRESSES: Please refer to Docket ID NRC-2018-0207 when contacting the 
NRC about the availability of information regarding this document. You 
may obtain publicly-available information related to this document 
using any of the following methods:
     Federal Rulemaking Website: Go to http://www.regulations.gov and search for Docket ID NRC-2018-0207. Address 
questions about Docket IDs in Regulations.gov to Jennifer Borges; 
telephone: 301-287-9127; email: [email protected]. For technical 
questions, contact the individual(s) 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 http://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``ADAMS Public Documents'' and 
then select ``Begin Web-based ADAMS Search.'' For problems with ADAMS, 
please contact the NRC's Public Document Room (PDR) reference staff at 
1-800-397-4209, 301-415-4737, or by email to [email protected]. The 
ADAMS accession number for each document referenced (if it is available 
in ADAMS) is provided the first time that it is mentioned in this 
document. In addition, for the convenience of the reader, the ADAMS 
accession numbers are provided in a table in the ``Availability of 
Documents'' section of this document.
     NRC's PDR: You may examine and purchase copies of public 
documents at the NRC's PDR, Room O1-F21, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852.

FOR FURTHER INFORMATION CONTACT: Christian Jacobs, Office of Nuclear 
Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001; telephone: 301-415-6825; email: 
[email protected].

SUPPLEMENTARY INFORMATION: 

I. Background

    Northern States Power Company-Minnesota, doing business as Xcel 
Energy (Xcel Energy, or the applicant) is the holder of Renewed 
Facility Operating License No. DPR-22, which authorizes operation of 
the MNGP, Unit No. 1, in Wright County, Minnesota, 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.
    Consistent with 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 ISFSI at power reactor sites 
to persons authorized to possess or operate nuclear power reactors 
under 10 CFR part 50. The applicant is authorized to operate a nuclear 
power reactor under 10 CFR part 50, and holds a 10 CFR part 72 general 
license for storage of spent fuel at the MNGP ISFSI. Under the terms of 
the general license, the applicant stores spent fuel at its ISFSI using 
the TN Americas LLC Standardized NUHOMS[supreg] dry cask storage system 
in accordance with CoC No. 1004, Amendments No. 9 and No. 10. As part 
of the dry storage system, the DSC (of which the closure welds are an 
integral part) ensures that the dry storage system can meet the 
functions of criticality safety, confinement boundary, shielding, 
structural support, and heat transfer.

II. Request/Action

    The applicant has requested 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 that require compliance with the 
terms, conditions, and specifications of CoC No. 1004, Amendment No. 
10, for the Standardized NUHOMS[supreg] Horizontal Modular Storage 
System, to allow continued storage of DSCs 11-15 in their respective 
Horizontal Storage Modules (HSMs). This would permit the continued 
storage of those five DSCs for the service life of the canisters. 
Specifically, the exemption would relieve the applicant from meeting TS 
1.2.5 of Attachment A of CoC No. 1004 (ADAMS Accession No. 
ML17338A114),

[[Page 47193]]

which requires that all DSC closure welds, except those subjected to 
full volumetric inspection, be dye penetrant tested in accordance with 
the requirements of the ASME B&PV Code Section III, Division 1, Article 
NB-5000. Technical Specification 1.2.5 further requires that the dye 
penetrant test (PT) acceptance standards be those described in 
Subsection NB-5350 of the ASME BP&V Code.
    Xcel Energy loaded spent nuclear fuel into six 61BTH DSCs starting 
in September 2013. Subsequent to the loading, it was discovered that 
certain elements of the PT examinations, which were performed on the 
DSCs to verify the acceptability of the closure welds, do not comply 
with the requirements of TS 1.2.5. All six DSCs were affected. Five of 
the six DSCs (numbers 11-15) had already been loaded in the HSMs when 
the discrepancies were discovered. DSC 16 remained on the reactor 
building refueling floor in a transfer cask (TC). On June 8, 2016, NRC 
granted an exemption (ADAMS Accession No. ML16159A227) 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 for DSC 16 only with regard to meeting 
TS 1.2.5 of Attachment A of CoC No.1004, Amendment No. 10. The 
exemption granted on June 8, 2016, restored DSC 16 to compliance with 
10 CFR part 72 and allowed Northern States Power Company-Minnesota to 
transfer DSC 16 into an HSM for continued storage at MNGP ISFSI for the 
service life of the canister.
    In a letter dated October 18, 2017 (ADAMS Accession No. 
ML17296A205) (Exemption Request), as supplemented in responses to NRC 
requests for additional information dated April 5, 2018 (ADAMS 
Accession No. ML18100A173) (RAI Response 1) and May 31, 2018 (ADAMS 
Accession No. ML18151A870) (RAI Response 2), the applicant requested an 
exemption from the following requirements to allow continued storage of 
the remaining DSCs 11-15 in their respective HSMs at the MNGP ISFSI:
     10 CFR 72.212(a)(2), which states that this general 
license is limited to storage of spent fuel in casks approved under the 
provisions of part 72;
     10 CFR 72.212(b)(3), which states that the general 
licensee must ensure that each cask used by the general licensee 
conforms to the terms, conditions, and specifications of a CoC or an 
amended CoC listed in 10 CFR 72.214;
     10 CFR 72.212(b)(5)(i), which requires that the general 
licensee perform written evaluations, before use and before applying 
the changes authorized by an amended CoC to a cask loaded under the 
initial CoC or an earlier amended CoC, which establish that the 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 10 CFR 72.214;
     10 CFR 72.212(b)(11), which states, in part, that the 
licensee shall comply with the terms, conditions, and specifications of 
the CoC and, for those casks to which the licensee has applied the 
changes of an amended CoC, the terms, conditions, and specifications of 
the amended CoC; and
     10 CFR 72.214, which lists the approved spent fuel storage 
casks.

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

    This exemption would permit the continued storage of DSCs 11-15 at 
the MNGP ISFSI for the service life of the canisters by relieving the 
applicant of the requirement to meet the PT requirements of TS 1.2.5 of 
Attachment A of CoC No. 1004. The provisions in 10 CFR part 72 from 
which the applicant is requesting exemption require the licensee to 
comply with the terms, conditions, and specifications of the CoC for 
the approved cask model it uses. Section 72.7 allows the NRC to grant 
exemptions from the requirements of 10 CFR part 72. As explained below, 
the proposed exemption will not endanger life or property, or the 
common defense and security, and is otherwise in the public interest. 
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 exemption is authorized by 
law.

Will Not Endanger Life or Property or the Common Defense and Security

    This exemption would relieve the applicant from meeting TS 1.2.5 of 
Attachment A of CoC No. 1004, which requires PT examinations to be 
performed on the DSCs to verify the acceptability of the closure welds, 
and would permit the continued storage of DSCs 11-15 in their 
respective HSMs at the MNGP ISFSI for the service life of the 
canisters. As detailed below, NRC staff reviewed the exemption request 
to determine whether granting of the exemption would cause potential 
for danger to life, property, or common defense and security.

Review of the Requested Exemption

    The NUHOMS[supreg] system provides horizontal dry storage of 
canisterized spent fuel assemblies in an HSM. The cask storage system 
components for NUHOMS[supreg] consist of a reinforced concrete HSM and 
a DSC vessel with an internal basket assembly that holds the spent fuel 
assemblies. The HSM is a low-profile, reinforced concrete structure 
designed to withstand all normal condition loads, as well as abnormal 
condition loads created by natural phenomena such as earthquakes and 
tornadoes. It is also designed to withstand design basis accident 
conditions. The Standardized NUHOMS[supreg] Horizontal Modular Storage 
System has been approved for storage of spent fuel under the conditions 
of CoC No. 1004. The DSCs under consideration for exemption were loaded 
under CoC No. 1004, Amendment No. 10.
    The NRC has previously approved the Standardized NUHOMS[supreg] 
Horizontal Modular Storage System. 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 exemption to assess the 
potential for danger to life or property or the common defense and 
security; the evaluation and resulting conclusions are presented below. 
The potential impacts identified for this exemption request were in the 
areas of materials, structural integrity, thermal, shielding, 
criticality, and confinement capability.
    Materials Review for the Requested Exemption: The applicant 
asserted that there is a reasonable assurance of safety to grant the 
requested exemption to continue the storage of DSCs 11-15 in their 
respective HSMs. The applicant's assertion of reasonable assurance of 
safety is based on the following factors:
     Reasonable assurance of weld integrity;
     Low dose consequences for a DSC in storage; and
     Low risk to the public.
    The applicant further stated that there is reasonable assurance of 
weld integrity based on the existing Quality Assurance (QA) 
documentation, engineering analysis, and expert evaluations, which 
demonstrate that the subject DSC welds

[[Page 47194]]

possess sufficient quality to perform their design functions due to the 
following:
     Fuel cladding integrity is maintained, as no damaged fuel 
was loaded and no unexpected dose readings were observed during drying 
operations.
     The weld design assures that there are no pinhole leaks 
and there is no credible process for service-induced flaws.
     The material, including the DSC shell, lids and weld 
filler, met quality requirements and quality welds were ensured by 
welding process qualification, welder qualification and the use of an 
automated welding process specifically designed for the application.
     In-process visual inspections of welds performed by the 
welders, Quality Control (QC) visual examination (VT) inspections of 
fit-ups and welds, and the vacuum hold, helium pressure and helium leak 
test all ensured confinement and quality of the welds.
     Strain margins for the DSC welds were demonstrated by 
structural analysis assuming flaw distributions conservatively derived 
from the Phased Array Ultrasonic Testing (PAUT) examination of DSC 16.
     Based on the DSCs 11-15 site-specific heat load 
conditions, additional margin exists to account for any remaining flaw 
uncertainty.
    The NRC materials review for the requested exemption focused on the 
applicant's assertion of reasonable assurance of weld integrity and 
each of the supporting assertions of: (1) Fuel cladding integrity; (2) 
weld design; (3) material and welding process; (4) tests performed; (5) 
adequate strain margins to accommodate flaws; and (6) additional strain 
margins in welds. A specific review of each of the supporting 
statements is provided in the following sections.
    Fuel Cladding Integrity: The applicant provided information on the 
nature of the spent nuclear fuel in DSCs 11-15 to demonstrate that the 
fuel cladding fission product barrier is intact and any postulated 
canister weld leak would have an insignificant effect on radioactive 
release. At the time of loading in 2013, the applicant stated that the 
combined decay heat load in the limiting DSC did not exceed 10.96 
kilowatts. In addition, only one of the 305 loaded fuel assemblies was 
considered to be high burnup, with a maximum recorded burnup of 45.12 
gigawatt days per metric ton of uranium (GWD/MTU) (in DSC 15). The 
applicant stated that cask loading reports and supporting 
radiochemistry records indicate that all of the fuel assemblies loaded 
into DSCs 11-15 met the TS requirements (TS Table 1-1t) for cladding 
integrity and no damaged fuel was loaded. The applicant stated that the 
integrity of the fuel was further demonstrated by the fact that no 
unexpected dose rate readings were observed during the vacuum drying 
processes of DSCs 11-15.
    The NRC staff reviewed the information provided by the applicant on 
the characteristics of the spent fuel loaded in DSCs 11-15. The NRC 
staff also reviewed the loading records for the loading campaign and 
confirmed that (1) no damaged fuel assemblies were loaded in the DSCs; 
(2) only one fuel assembly had burnup that marginally exceeded the 45 
GWD/MTU criterion for high burnup fuel however, the cladding of the 
fuel assembly was shown to be intact through cask loading reports and 
supporting radiochemistry reports; and (3) no unexpected dose readings 
were observed in the loading campaign. Based on the review of the 
information from the loading campaign, the NRC staff confirmed that the 
characteristics of the fuel loaded in the DSCs included in the 
exemption request were accurately described.
    Weld Design: The applicant stated that the updated final safety 
analysis report (UFSAR) only describes weld failure in terms of a 
possible pinhole leak in individual weld layers. The applicant further 
stated that the UFSAR assumes or stipulates that pinholes may exist in 
individual layers but the UFSAR makes no explicit mention about how a 
pinhole leak in a weld layer is formed, whether it occurs during the 
weld formation or by subsequent canister loading operations, fatigue 
cycles during storage, or accidents. The applicant stated that the 
existence of pinhole leaks is a non-mechanistic assumption of the 
UFSAR; and there is no underlying malfunction that causes its 
formation.
    The applicant stated that, once in storage, there is no credible 
failure mechanism of the DSC top cover plate closure welds that would 
adversely affect DSC confinement because (1) the top cover plate and 
weld material are stainless steel and the only welds subject to the 
outside environment are the outer layer of the outer top cover plate 
(OTCP) weld and the test port plug (TPP) weld; (2) a reduction in cross 
section from plastic strain is not applicable to the top cover plate 
welds because the differential pressure across the top cover plates 
conditions is minimal (less than one atmosphere); and (3) the mechanism 
of cyclic loading is not applicable to the top cover plate and closure 
welds because the extent of fatigue cycling experienced by the canister 
is below the threshold which the ASME B&PV Code Section III has 
established.
    The NRC staff have previously reviewed the design of the 
NUHOMS[supreg] 61BTH DSC included in the UFSAR. The NRC staff verified 
that the top cover plate and weld material are stainless steel and the 
only welds subject to the outside environment are the outer layer of 
the OTCP weld and the TPP weld. The NRC staff verified that the 
differential pressure across the top cover plates is minimal and 
consequently the reduction in cross section from plastic strain is not 
credible. The NRC staff have reviewed the assessment of fatigue and 
determined that the DSCs are not subjected to cyclic loading that 
requires a fatigue analysis. Based on the NRC staff's previous analysis 
of the DSC weld design, the NRC staff determined that the applicant's 
assessment of the weld design is accurate and there is no credible 
mechanism for the propagation of an existing weld flaw to result in a 
through weld thickness penetration that would result in a leak.
    Material and Welding Process: The applicant stated that procurement 
records such as certified material test reports (CMTRs) demonstrate 
that the canisters, lids, and weld filler materials met design 
standards and quality requirements, thereby assuring compatibility 
between materials and satisfactory material performance characteristics 
(e.g., material strength).
    The applicant stated that the weld closures of DSCs 11-15 were 
performed under a 10 CFR part 50 Appendix B QA program, such that the 
canister integrity is assured. The applicant stated that welding 
materials were procured to quality requirements, welding processes were 
developed and qualified for the given configuration, and welders were 
appropriately qualified to the ASME B&PV Code requirements. Finally, 
the applicant stated that welding parameters were specified in 
associated procedures and monitored as required.
    In addition to the original weld head video review conducted in 
conjunction with the DSC 16 exemption request, the applicant included 
another examination of the weld head video and the general area videos 
taken during the 2013 cask loading campaign. Based on the examination 
of the videos, the applicant made a correlation between weld techniques 
and typical weld flaw characteristics such as those identified in the 
PAUT of the inner top cover plate (ITCP) and OTCP welds from DSC 16. 
The applicant provided an assessment conducted by Structural Integrity

[[Page 47195]]

Associates, Inc. (SIA), which concluded that defects would be limited 
in the through thickness dimension to the thickness of a single bead. 
The applicant also stated that, even considering the possibility that 
any given layer of weld may have a leak through that layer, the 
licensing basis criterion stated in the UFSAR Section 3.3.2.1 assures 
that the chance of pinholes being in alignment on successive 
independently-deposited weld layers is not credible.
    As stated above, the NRC staff have previously reviewed the design 
of the NUHOMS[supreg] 61BTH DSC included in the UFSAR. The NRC staff 
reviewed the materials used in the construction of DSCs 11-15 and the 
NRC staff confirmed that the materials used met the specifications 
called out in the NUHOMS[supreg] 61BTH DSC design. The NRC staff 
reviewed the CMTRs and confirmed that the materials met specified 
compositional and mechanical property requirements.
    The NRC staff reviewed, ``TRIVIS Inc. Welding Procedure 
Specification (WPS) SS-8-M-TN, Revision 10,'' (Enclosure 2 to RAI 
Response 1) which was used for the machine welding of the ITCP and the 
OTCP as well as, ``TRIVIS Inc. WPS SS-8-A-TN, Revision 8,'' (RAI 
Response 1 Enclosure 3) used for manual welding of the ITCP and the 
OTCP. The NRC staff compared WPS SS-8-M-TN, Revision 10 and WPS SS-8-A-
TN, Revision 8 to the essential variables required for the gas tungsten 
arc welding (GTAW) in ASME Section IX Part QW Welding, Article II 
Welding Procedure Qualifications, Table QW-256 and Article IV Welding 
Data, Subsection QW-400 Variables. The NRC staff determined that the 
WPS SS-8-M-TN, Revision 10 and WPS SS-8-A-TN, Revision 8 are acceptable 
because all of the essential variables identified in ASME Section IX 
for GTAW WPSs were included and the range of permissible values were 
specified.
    The NRC staff reviewed, ``TRIVIS, Inc. Procedure Qualification 
Record (PQR) PQR-1, Revision 2'' (Enclosure 4 to RAI Response 1). The 
NRC staff compared the testing documented in PQR-1, Revision 2 against 
ASME Section IX Part QW Welding, Article I Welding General 
Requirements. The NRC staff determined that PQR-1 Revision 2 was 
acceptable because all the testing necessary to qualify WPS SS-8-M-TN, 
Revision 10 and WPS SS-8-A-TN, Revision 8 were performed with 
satisfactory results and documented in PQR-1, Revision 2.
    As documented in NUREG-1536, Revision 1, Section 8.9.1 (ADAMS 
Accession No. ML101040620) the NRC previously determined that for a 
multipass lid-to-shell weld of an austenitic stainless steel canister 
designed and fabricated in accordance with the ASME B&PV Code Section 
III Subsection NB (Class 1 components), no flaws of significant size 
will exist such that the flaws could impair the structural strength or 
confinement capability of the weld. For a spent nuclear fuel canister, 
such a flaw would be the result of improper fabrication or welding 
technique, as service-induced flaws under normal and off-normal 
conditions of storage are not credible.
    The NRC staff notes that per the guidance in NUREG-1536, Revision 
1, Section 8.4.7.4, the large structural lid-to-shell weld designs 
fabricated from austenitic materials may be tested using non-
destructive examination methods such as a volumetric ultrasonic test 
(UT) or a multi-pass PT. If a multiple-pass PT examination is utilized 
in lieu of UT inspection, a stress reduction factor of 0.8 for weld 
strength is imposed. In the absence of valid PT examinations of the 
closure welds for DSCs 11-15, the applicant asserted that the helium 
leak rate tests performed on all DSCs and the PAUT results for DSC 16, 
which show that weld defects are limited to the height of one weld 
bead, support the claim that DSCs 11-15 do not have flaws that would 
impair the structural strength or confinement capability.
    The NRC staff reviewed the information provided by the applicant 
including the DSC lid-to-shell closure weld design for the ITCP and the 
OTCP, the manual and machine GTAW WPSs, the helium leak testing results 
for DSCs 11-15 and the PAUT results for DSC 16. The NRC staff concluded 
that the design of the DSC closure weld and the GTAW WPSs used to weld 
the ITCP and the OTCP are unlikely to result in weld flaws that could 
impair the structural strength or confinement capability of the weld. 
The NRC staff concluded that the helium leak testing results for DSCs 
11-15 confirmed that there were no flaws that impaired the confinement 
capability of the DSC 11-15 ITCP welds. The NRC staff concluded that 
the PAUT results for DSC 16 is sufficient to show that the GTAW of the 
ITCP and OTCP welds do not result in defects that would impair 
structural strength or confinement capability of the DSC closure welds.
    Tests Performed: The applicant stated that a number of independent 
tests were conducted on the DSC 11-15 welds which verify that adequate 
welds were performed on DSCs 11-15. The applicant stated that these 
tests include:
     In-process visual examination and QC visual examinations 
to demonstrate that weld processes were followed and a weld meeting 
visual examination criteria was developed; and
     Helium leakage tests to verify the confinement integrity 
function and, to some extent, the structural integrity function of the 
DSC welds.
    The applicant provided an extent of condition assessment as 
Appendix D of Enclosure 1 of the Exemption Request. The applicant 
stated that the extent of condition assessment was focused on:
     Compliance with welding administrative requirements;
     Technical specification required testing of welds; and
     Weld depth measurements for outer top cover plate welds.
    The NRC staff reviewed the information provided in the application 
and confirmed that the applicant provided documentation that the 
welding administrative requirements were met, as follows: (1) Welding 
procedures were available at the job site for welding operators to 
follow; (2) weld surface preparations were completed such that the weld 
surface was dry and free of oil, grease, weld spatter, rust, slag, 
sand, discontinuities, or other extraneous material; (3) weld crown 
height for the ITCP and vent/siphon port were verified; and (4) welds 
for the ITCP, OTCP and the vent and siphon ports were all verified.
    The NRC staff reviewed the information provided in the application 
and confirmed that the applicant provided documentation for the TS 
required tests performed on DSCs 11-15. The NRC staff verified that the 
application included documentation showing that (1) hydrogen monitoring 
was properly performed while welding in accordance with TS 1.1.11; (2) 
pressure testing of the DSC shell to ITCP weld was conducted in 
accordance with TS 1.1.12.4; (3) two cycles of vacuum drying and 
verification were conducted at a vacuum less than 2.8 torr and were 
maintained for times longer than 30 minutes in accordance with TS 
1.2.2; (4) the DSCs were backfilled with helium and to a pressure of 
17.2  1.0 psi for a time of at least 30 minutes in 
accordance with TS 1.2.3a; and (5) helium backfilling, pressure 
verification and leak testing were conducted in accordance with 
American National Standards Institute (ANSI) N14.5-1997 and leak rates 
less than 1.0 x 10-7 ref cubic centimeters/sec were 
documented for DSCs 11-15 in accordance with TS 1.2.4a.
    The NRC staff confirmed that the weld depth measurements for the 
OTCP were conducted at four locations around the weld circumference. 
The NRC staff confirmed that the weld depth (dimension of the weld 
throat)

[[Page 47196]]

measurements met the minimum requirements of 0.5 inches for the OTCP 
weld for DSCs 11-15.
    Based on the review of the information provided by the applicant, 
the NRC staff determined that the required tests were performed on the 
ITCP and OTCP welds including in-process visual inspections of welds 
performed by the welders, VT of fit-ups and welds and the vacuum hold, 
as well as helium pressure and helium leak testing. The NRC staff 
determined that the applicant completed an adequate extent of condition 
assessment which showed that the welding of the ITCP and OTCP were 
conducted in accordance with welding administrative requirements, the 
required testing of welds were in compliance with technical 
specifications, and weld depth measurements for the OTCP met design 
requirements for the 61BTH DSC. Adequate Strain Margins to Accommodate 
Flaws (Exemption Request Enclosures 2 through 5): The applicant stated 
that strain margins for DSCs 11-15 were demonstrated by structural 
analysis using theoretically-bounding full-circumferential flaws and a 
structural analysis assuming flaw distributions conservatively derived 
from the PAUT examination of DSC 16. The applicant supported the 
analysis using:
     A review of weld head video for all available DSCs, 
general area video for all available DSCs, and welding records;
     the allowable flaw size evaluation in the ITCP closure 
weld for DSC 16; and
     the ITCP and OTCP closure weld flaw evaluation for a 61BTH 
DSC based on the DSC 16 PAUT results.
    Based on the review of the videos, welding records and the PAUT 
examination of DSC 16, the applicant determined that the indications 
found on DSC 16 are representative of those that may be found on DSCs 
11-15. Consequently, the applicant determined that the same bounding 
analyses performed for DSC 16 should provide for similar conservative 
results for the closure welds for DSCs 11-15. The applicant stated that 
for the OTCP, the original design basis calculations determined 
critical flaw sizes. The applicant stated that these design basis 
analyses determined for a 360[deg] circumferential flaw, an allowable 
flaw depth of 0.19 inch and 0.29 inch could exist for surface connected 
and sub-surface flaws respectively. Finally, the applicant stated that 
the flaw sizes determined by these calculations bound any of the 
indications found on DSC 16 by PAUT of the OTCP weld.
    For the ITCP weld of DSC 16, the applicant provided a calculation, 
AREVA Calculation 11042-0204, Revision 3, ``Allowable Flaw Size 
Evaluation in the Inner Top Cover Plate Closure Weld for DSC #16'' 
(Exemption Request Enclosure 4) that documents the critical flaw size 
based on the maximum radial stresses in the welds due to design loads. 
The applicant's analysis calculated the critical flaw size for a weld 
size of 0.25 inch per the PAUT results for DSC 16, which showed that 
the distance between the weld root and crown at the canister wall for 
the DSC 16 ITCP lid weld ranged from 0.25 inch to 0.4 inch. The 
applicant determined that the critical flaw depth was 0.15 inch, which 
would exceed the typical weld layer thickness. The applicant noted that 
the measured weld size for the ITCP weld on DSC 16 was significantly 
larger than the design thickness of 3/16 inch (i.e., 0.188''). The 
applicant stated that all analyses for DSCs 11-15 were conducted using 
the design thickness of the weld. The applicant provided an analysis of 
the allowable flaw size for the DSC ITCP and OTCP using the weld design 
thickness which used the flaw sizes from the PAUT examination of DSC-16 
(Exemption Request Enclosure 5, AREVA Calculation 11042-0205, Revision 
3, ``61BTH ITCP and OTCP Closure Weld Flaw Evaluation'').
    The applicant stated that, as part of the original extent of 
condition review, weld head videos were reviewed by SIA in 2014. For 
DSCs 13 and 16, the review included video recordings of the ITCP root 
and cover weld layers and the OTCP tack, root, intermediate and cover 
weld layers. For DSCs 12, 14 and 15, the review included video 
recordings of the OTCP tack, root, intermediate and cover weld layers. 
The applicant stated that no weld head video was available for DSC 11. 
The DSC 16 outer closure weld was concluded to be the most vulnerable 
to potential defects because a greater frequency of irregular surface 
conditions was generated during welding.
    The applicant stated that SIA performed further reviews of 
available weld head videos along with general area videos, welding 
records, and PAUT results for DSC 16 to identify any correlations 
between the welding processes used during the 2013 loading campaign and 
the flaws identified by the PAUT. The applicant stated that, by 
correlating indications to the particular welding methods used on all 
six canisters (including DSCs 11-15), a reasonable case was made that 
the types of indications found on DSC 16 are representative of those 
that may be found on DSCs 11-15.
    For the OTCP, the applicant stated SIA concluded that the defects 
located within the weld deposit of DSC 16 are believed to be inter-bead 
lack of fusion formed at the interface between adjacent weld bead 
surfaces. The applicant stated that when the defects are present in the 
DSC OTCP closure weld, they would be found at the interfaces between 
weld beads. The applicant included a schematic showing the DSC OTCP 
weld bead placement and the position of the lack-of-fusion flaws, which 
were characterized as parallel and offset. The applicant stated that 
the possible locations where lack of fusion between the sides of 
adjacent weld beads could form in the DSC OTCP closure weld would 
result in defects that are not aligned and which would not extend 
beyond the thickness of one weld pass layer.
    For the ITCP, the applicant stated SIA concluded that the locations 
of the flaws in DSC 16 indicate that they were related to sidewall lack 
of fusion. SIA also noted that the weld joint geometry, welding system, 
and welding setup for the ITCP of DSCs 11-15 had potential for forming 
defects on the sidewall like those identified in DSC 16. The applicant 
stated that, from the review, SIA concluded the other five canister 
ITCP closure welds were welded in a similar manner, using similar 
welding procedures, equipment, welding process, filler material, and 
welding operators and thus, it is reasonable to assume the other 
canister ITCP welds will have similar intermittent defects. In 
addition, the applicant stated that the vertical weld wall of the weld 
groove is inherent to a single bevel design, and because there is 
limited room to tilt the tungsten electrode towards the side wall (DSC 
shell), any lack-of-fusion defects that might form would likely be 
located on the vertical sidewall. The applicant concluded that the 
assumptions made for the ITCP closure weld bounding analysis in DSC 16 
were considered reasonable for all ITCP canister closure welds.
    The NRC staff reviewed the applicant's summary of the weld head 
video and general area videos. The NRC staff also reviewed the 
applicant's supporting analyses including:
     AREVA Calculation 11042-0204, Revision 3, ``Allowable Flaw 
Size Evaluation in the Inner Top Cover Plate Closure Weld for DSC #16'' 
(Exemption Request Enclosure 4);
     AREVA Calculation 11042-0205, Revision 3, ``61BTH ITCP and 
OTCP Closure Weld Flaw Evaluation'' (Exemption Request Enclosure 5);

[[Page 47197]]

     Structural Integrity Associates, Inc. Report 700388.401, 
Revision 1, ``Evaluation of the Welds on DSC 11-15'' (Exemption Request 
Enclosure 3);
     Structural Integrity Associates Inc. Report 1301415.403, 
Revision 2, ``Assessment of Monticello Spent Fuel Canister Closure 
Plate Welds Based on Welding Video Records'' dated May 22, 2014 (RAI 
Response 1 Enclosure 8);
     Structural Integrity Associates Inc. Report 1301415.402, 
Revision 0, ``Review of TRIVIS Inc. Welding Procedures used for Field 
Welds on The Transnuclear NUHOMS[supreg] 61BTH Type 1 & 2 Transportable 
Canister for BWR Fuel'' (RAI Response 1 Enclosure 9); and
     RAI Response 2.
    The NRC staff determined that, because the same welding process, 
welding equipment, and welding procedures were used by the personnel 
that conducted the ITCP and OTCP welds in DSCs 11-16, it is reasonable 
to conclude, based on engineering judgement that the types of defects 
in DSC 16 are representative of those that may be in DSCs 11-15. The 
NRC staff determined that, because the DSCs 11-16 are the same design, 
were fabricated to the same specifications, and were subjected to the 
same tests, the analysis conducted for DSC 16 is also applicable to 
DSCs 11-15.
    The NRC staff reviewed the applicant's analysis for the OTCP welds 
and the description of the OTCP welding based on weld head video 
described in Exemption Request Enclosure 3, Structural Integrity 
Associates, Inc. Report 700388.401, Revision 1, ``Evaluation of the 
Welds on DSC 11-15,'' Appendix B, ``Outer Top Cover Plate Closure Weld 
Bead Sequence (Based on VID Observations)'' and Appendix C, ``Tabulated 
Review of Available VIDS for Monticello DSC-12 thru DSC-16.'' The NRC 
staff also reviewed the information included from the review of the 
general area video records included in Appendix D of Exemption Request 
Enclosure 3, ``Monticello DSC Video Inspection.'' The NRC staff 
determined that due to the OTCP weld joint design and welding process 
used in the OTCP closure weld, the likely significant welding defects 
in the OTCP weld would be lack of fusion between the weld beads or at 
the interface of the OTCP weld and the OTCP or the interface of the 
OTCP weld and the DSC shell. Given the geometry of the weld joint, the 
number of welding passes required to fill the weld joint, the position 
of each welding pass, and the requirement for in-process visual 
inspection of the weld after each pass, the NRC staff determined that 
it is unlikely that a connected lack-of-fusion defect greater than the 
thickness of one pass would be present. The NRC staff determined that 
any lack-of-fusion defects in the OTCP would not be aligned because of 
the weld joint geometry and the positioning of the weld passes required 
to fill the OTCP weld joint.
    With respect to the ITCP welds, the NRC staff reviewed the 
applicant's analysis for the ITCP welds and the description of the ITCP 
welding based on weld head video described in Exemption Request 
Enclosure 3, Structural Integrity Associates, Inc. Report 700388.401, 
Revision 1, ``Evaluation of the Welds on DSC 11-15.'' The NRC staff 
also reviewed the following appendices to Exemption Request Enclosure 
3: Appendix A, ``Inner Top Cover Plate Closure Weld Bead Sequence 
(Based on VID Observations)''; Appendix C, ``Tabulated Review of 
Available VIDS for Monticello DSC-12 through DSC-16''; and Appendix D 
``Monticello DSC Video Inspection.''
    The NRC staff notes that it is unclear whether some of the 
observations in Exemption Request Enclosure 3, Appendix C were in 
conformance with Procedure 12751-MNGP-OPS-01, Revision 0, ``Spent Fuel 
Cask Welding: 61BT/BTH NUHOMS[supreg] Canisters'' (RAI Response 1 
Enclosure 6). In particular, the NRC staff note that Exemption Request 
Enclosure 3, Appendix C indicated there were two instances of blow 
through of the root pass on the OTCP weld of DSC-12. Procedure 12751-
MNGP-OPS-01, Revision 0 states such an event would be treated as a 
major repair with additional NDE and documentation. However, in RAI 
Response 2, the applicant indicated that these events were weld craters 
and were not weld root blow through events. While NRC staff was not 
able to resolve whether these actions taken by the welder were in 
conformance with the applicable procedure, it was apparent from 
Exemption Request Enclosure 3, Appendix C that corrective actions were 
taken to address the weld defects. In addition, the NRC staff 
determined that either a blow through of the root pass or a weld crater 
is a localized defect that would, in the worst case, compromise a small 
length of the root pass. As such, the NRC staff determined that the 
reported observation of a possible root blow through in two locations 
is bound by the assumed size of the OTCP welds defects in the flaw 
evaluation.
    The NRC staff determined that for the ITCP weld joint design the 
likely significant welding defects would be lack of fusion at the 
interface of the ITCP weld and the ITCP or the interface of the ITCP 
weld and the DSC shell. Given the geometry of the weld joint, the 
number of welding passes required to fill the weld joint, the position 
of each welding pass, and the requirement for in-process visual 
inspection of the weld after each pass, the NRC staff determined that 
lack of fusion between the ITCP weld and the DSC shell is likely to be 
the most significant type of weld defect in this joint. The NRC staff 
determined that the positioning of the welding electrode necessary to 
weld the root pass would minimize the chances of a lack-of-fusion 
defect located at the interface of the ITCP weld and the ITCP. The NRC 
staff determined that the positioning of the welding electrode 
necessary to weld the second fill pass would minimize the chances of a 
lack-of-fusion defect at the interface of the ITCP weld and the DSC 
shell.
    Based on the review of the information provided by the applicant 
including the review of weld head video for all available DSCs, general 
area video for all available DSCs, and welding records; the allowable 
flaw size evaluation in the ITCP closure weld for DSC 16; and the ITCP 
and OTCP closure weld flaw evaluation for a 61BTH DSC based on the DSC 
16 PAUT results, the NRC staff concludes that the applicant has 
adequately considered the sizes and location of potential weld flaws to 
evaluate the stress margins in the ITCP and OTCP welds of DSCs 11-15. 
The NRC staff structural review for the requested exemption follows the 
materials review.
    Additional Strain Margins in Welds (Exemption Request Enclosures 6 
through 9): The applicant stated that additional analysis was performed 
to maximize the size of flaws present in locations consistent with the 
results of the DSC 16 PAUT to demonstrate substantial margin to account 
for potential flaw uncertainties. In addition, the applicant stated 
that DSCs 11-15 site-specific heat load conditions were applied to 
demonstrate additional weld margin exists and is available to account 
for any remaining flaw uncertainty. The applicant stated that the 
analysis used design basis loads with flaws present in locations 
consistent with the DSC 16 PAUT results and maximized in size such that 
the weld flaws approach acceptable design limits.
    The applicant stated that the two maximum modeled weld flaws for 
OTCP to DSC shell weld are 0.43 inch and 0.42 inch in height, which 
represents about 85% through-wall of the 0.5-inch minimum weld throat. 
The applicant stated that the maximum modeled full-circumferential weld 
flaws

[[Page 47198]]

for ITCP to DSC shell weld are 0.11 inch in height at the ITCP weld to 
the ITCP interface and 0.14 inch in height at the ITCP weld to DSC 
shell interface, which represent respectively 58% and 74% through-wall 
of the 0.19-inch minimum weld throat. The applicant stated that each of 
the four assumed flaws represent defects spreading over more than one 
weld bead.
    The NRC staff reviewed the applicant's analysis for the ITCP and 
OTCP weld flaws along with the applicant's summary of the welding video 
recordings and the PAUT examination results for DSC 16. For the ITCP 
weld, the NRC staff assessed the geometry of the weld joint, the 
positioning of the welding electrode in both the root and the final 
fill pass along with the requirement for in-process visual inspection 
of the weld after each pass. For the OTCP weld, the NRC staff assessed 
the geometry of the weld joint, the number of welding passes required 
to fill the weld joint, the position of each welding pass, along with 
the requirement for in-process visual inspection of the weld after each 
pass. The NRC staff determined that any lack-of-fusion defects in the 
ITCP and OTCP would not be aligned and would not result in a defect 
greater than the thickness of one pass given the weld joint geometry 
and the positioning of the weld passes required to fill the ITCP and 
OTCP weld joints. Thus, the NRC staff determined that the flaws 
assessed in Exemption Request Enclosure 6 are both unlikely to occur in 
any of the DSCs loaded in the 2013 campaign and the flaws assessed in 
Exemption Request Enclosure 6 conservatively bound any possible welding 
defects that are likely to exist in the DSC 11-15 OTCP welds.
    Based on the review of the information provided by the applicant 
including the analysis of flaws analyzed from the PAUT examination of 
the ITCP and OTCP welds of DSC 16 and the assumed maximized flaws that 
exceed the weld bead deposit thickness, the NRC staff concludes that 
the applicant's analysis of stress margins in the ITCP and OTCP welds 
of DSCs 11-15 conservatively assumed weld flaws that are much larger 
than would be reasonably expected. This is due to the combination of 
the materials of construction, weld joint designs, and the welding 
process used for the ITCP and OTCP welds.
    Structural Review for the Requested Exemption: The exemption 
request states that there is a reasonable assurance of safety to grant 
the requested exemption to continue the storage of DSCs in their 
respective HSMs. As noted by the applicant, one of the many factors 
contributing to this assertion is the structural integrity of the DSC 
top cover plates-to-shell closure welds. The Structural Review is based 
on the conclusion of the Materials Review where the NRC staff 
determined among other findings that, because the DSCs 11-16 are of the 
same design, were fabricated to the same specifications, and were 
subjected to the same tests, the analyses conducted for DSC 16 may also 
be applied to DSCs 11-15.
    For the DSC 11-15 closure weld structural functions assessment, 
which was done by analysis, the applicant noted that the previous 
evaluations to demonstrate adequate strain margins of safety of the DSC 
16 closure welds also support the current exemption request. These 
evaluations were provided in the following reports:
     SIA Report 1301415.301, Revision 0, ``Development of an 
Analysis Based Stress Allowable Reduction Factor (SARF)--Dry Shielded 
Canister (DSC) Top Closure Weldments'' (Exemption Request Enclosure 2);
     AREVA Calculation 11042-0204, Revision 3, ``Allowable Flaw 
Size Evaluation in the Inner Top Cover Plate Closure Weld for DSC #16'' 
(Exemption Request Enclosure 4); and
     AREVA Calculation 11042-0205, Revision 3, ``61BTH ITCP and 
OTCP Closure Weld Flaw Evaluation'' (Exemption Request Enclosure 5).
    The evaluations performed on the DSC 16 closure welds included: (1) 
A structural analysis using an analysis-based stress allowance 
reduction factor and theoretically-bounding full-circumferential flaws 
to demonstrate that finite element analysis (FEA) simulation is 
suitable for analyzing the structural performance of the weld as a 
continuum with multiple embedded flaws; (2) a calculation that 
documents the allowable critical flaw size in the ITCP closure weld 
based on the maximum design basis radial stresses in the welds; and (3) 
a structural analysis demonstrating large weld strain margins of safety 
with conservative assumptions of flaw distribution and size derived 
from the DSC 16 PAUT examination results.
    However, to demonstrate adequate strain margin and to accommodate 
flaws in the DSCs 11-15 closure welds, the applicant provides a FEA 
simulation evaluation in SIA Report, 700388.401, Revision 1, 
``Evaluation of the Welds on DSCs 11-15,'' (Exemption Request Enclosure 
3) to support that the flaw distribution and size based on the PAUT 
examination results for the DSC 16 closure weld performance can be used 
to conservatively represent the closure weld flaws for DSCs 11-15. As 
noted in the Materials Review, the NRC staff reviewed the applicant's 
evaluation and determined that the flaws used in analyzing the DSC 16 
closure welds are a reasonable representation for the closure welds for 
all DSCs 11-16. This finding provides the basis for the NRC staff to 
review the two calculation packages: Calculations 11042-0207 and 11042-
0208, which used the maximized weld flaws that are essentially the same 
in distribution but are much larger in size than those used for the DSC 
16 evaluation.
    Specifically, in Calculation 11042-0207, the applicant asserts that 
there are adequate strain margins in the welds to accommodate flaws for 
DSCs 11-15. The DSCs are subject to the design basis temperature, 
pressure, and side-drop loading conditions and are analyzed per the 
ASME Code Section III criteria, using the limit load and elastic-
plastic analyses. In Calculation 11042-0208, the applicant asserts 
additional strain margin in the DSCs 11-15 closure welds. The maximum 
flaws, the analysis methodology and the evaluation criteria are the 
same as those of Calculation 11042-0207. However, in lieu of the design 
basis loading, the analysis used the as-loaded DSC cavity pressure, 
which is site-specific and temperature dependent. The at-temperature 
material yield strengths are used, which are higher than those 
associated with the design basis loading.
    It is noted that the exemption request also included Calculation 
11042-0209 (Exemption Request Enclosure 8) to demonstrate additional 
weld strain margin for DSCs 11-15 subject to the site-specific side-
drop loading condition. The NRC staff neither approves, nor rejects, 
and is not expressing any view related to the material in the 
calculation, as it did not enter into the NRC evaluation.
    The NRC staff reviewed the above two calculation reports on the 
structural performance of the DSC 11-15 closure welds. In Calculation 
11042-0207, the applicant followed the same analysis method used in 
Calculation 11042-0205 for DSC 16 to demonstrate adequate strain margin 
in DSCs 11-15 closure welds. The applicant noted that the finite 
element model details and structural performance acceptance criteria 
are the same except that the maximized flaw configuration is postulated 
to result in much larger flaws than those associated with DSC 16 to 
provide additional insights into the weld structural performance.
    To arrive at the maximized configuration, the flaws modeled in

[[Page 47199]]

Calculation 11042-0205 for DSC 16 were first modified slightly, 
including replacing conservatively the 0.11 inch-long flaw inside the 
ITCP with an equivalent-height flaw at the interface between the ITCP 
and the 3/16-inch ITCP-to-shell weld. However, the size and location of 
all other welds were unchanged. Next, an elastic-plastic analysis of 
flaw length introduced increasingly larger flaw sizes in each analysis 
iteration to simulate higher localized plastic strain. As noted by the 
applicant, the iteration analysis was considered complete for the 
maximized flaws determination for which the peak equivalent plastic 
strain for the most critically stressed flaws would be calculated to be 
somewhat below the ASME code weld material elongation limit of 28 
percent. The applicant performed the elastic-plastic iteration analysis 
using a 150-percent design basis side-drop of 112.5 g (75 x 1.5 = 
112.5) to arrive at the maximized flaws. Specifically, the maximized, 
360[deg] full-circumferential flaws are of 0.43 inch and 0.42 inch in 
height for the two flaws associated with the OTCP, which represent 
about 85% through-wall of the 0.5-inch minimum throat for OTCP-to-DSC 
shell weld. The maximized full-circumferential flaws for ITCP-to-DSC 
shell weld are 0.11 inch and 0.14 inch each in height, which represent 
respectively 58% and 74% through-wall of the 0.19-inch minimum weld 
throat. The NRC staff reviewed the iteration analysis for arriving at 
the maximized flaws for the DSCs 11-15 closure welds. Because the 
maximized flaws are essentially the same in locations as those used for 
DSC 16 and the resulting flaw sizes are much larger than the 
corresponding ones used for DSC 16, the NRC staff concludes that the 
postulated maximized flaws are conservative and appropriate for 
evaluating the strain performance of the DSCs 11-15 closure welds.
    Using the maximized flaws, the applicant performed limit load 
analyses in Calculation 11042-0207 for two DSC design basis internal 
pressures of 32 psi and 65 psi for the ASME Code Service Level A/B and 
Service Level D evaluations, respectively. The analyses resulted in the 
calculated collapse pressures of 86.3 psi for Service Level A/B and 
122.2 psi for Service Level D. The collapse pressures are acceptable 
because they are greater than the respective ASME Code limit-load 
analysis acceptance criteria of 60 psi and 90.2 psi. Similarly, for the 
design basis DSC side-drop of 75 g, the applicant used the 3D half-
symmetric model to perform a Service Level D limit load analysis. The 
applicant determined the side-drop collapse load to be approximately 
179.5 g, which includes an off-normal DSC design basis internal 
pressure of 20 psi as a boundary condition. This determination is 
acceptable because the collapse load is greater than the required side-
drop load of 104 g to satisfy the ASME Code limit-load analysis 
acceptance criteria.
    To address the potential material rupture associated with high 
plastic strain concentrations at the weld flaws, the applicant 
performed elastic-plastic analyses in Calculation 11042-0207 to 
quantify strain margins of safety for the DSCs 11-15 with maximized 
flaws. This concern was addressed by considering a Ramberg-Osgood 
idealization of the stress-strain curve for SA-240 Type 301 stainless 
steel, which recognizes strain hardening effects for the FEA modeling. 
The elastic-plastic analyses resulted in the peak equivalent plastic 
strains of 7.4 percent and 11.1 percent for the Service Level D design 
basis pressure of 65 psi and side-drop of 75 g, respectively. For the 
strain margin evaluation, the applicant continued to use the same DSC 
16 weld strain acceptance criterion of not exceeding the 28 percent 
elongation limit, which is a reduction from the ASME B&PV Code 
specified weld elongation limit of 35 percent by a factor of 0.8 (0.35 
x 0.8 = 0.28). Considering the 28 percent elongation limit, the strain 
margins of safety corresponding to the calculated peak equivalent 
plastic strains are 2.78 {(0.28/0.074)-1 = 2.78{time}  and 1.52 {(0.28/
0.111)-1 = 1.52{time} , respectively. Because the margins of safety are 
all positive (i.e., greater than zero), the NRC staff concludes that 
there are adequate strain margins in the welds to accommodate flaws for 
DSCs 11-15.
    Additionally, similar to the analysis used to supplement 
qualification of the DSC 16 closure welds, the applicant considered a 
150 percent of the design basis loading to evaluate the DSCs 11-15 
welds. The analysis used a DSC internal pressure of 100 psi (65 x 1.5 = 
97.5 <100 psi) and a side-drop of 112.5 g (75 x 1.5 = 112.5 g), which 
are beyond the ASME B&PV Code, Section III, Paragraph NB-3228.3 Plastic 
Analysis provisions. The calculated peak equivalent plastic strains are 
13.6 percent and 23.0 percent for the respective pressure and side-drop 
loading cases. For the weld strain margin evaluation, the applicant 
continued to use the same 28 percent weld elongation limit which 
resulted in the weld strain margins of safety of 1.06 {(0.28/0.0136)-1 
= 1.06{time}  and 0.22 {(0.28/0.23)-1 = 0.22{time} , respectively. 
Because all margins of safety are positive, even in loading conditions 
that are 50 percent beyond those required for evaluating localized 
strains by the elastic-plastic analysis, the NRC staff concludes that 
there are adequate strain margins on the welds to accommodate flaws for 
DSCs 11-15.
    The applicant noted that there are additional strain margins in the 
closure welds of DSCs 11-15 owing to the site-specific as-loaded 
temperature and DSC internal pressure conditions at MNGP, which are 
less severe than those associated with the design basis conditions. In 
Calculation 11042-0208 (Exemption Request Enclosure 7), the applicant 
performed evaluations using the temperature and pressure conditions 
specific to DSCs 11-15. The evaluation follows the same Calculation 
11042-0207 analysis method and acceptance criteria, including the same 
maximized flaws. The applicant indicated that the evaluations were 
intended to address any remaining uncertainties related to potential 
flaws that may be present in DSCs 11-15 by demonstrating existence of 
additional strain margins in the closure welds.
    Using the site-specific 370 [deg]F at-temperature material yield 
strength of 21.2 ksi for the SA-240 Type 304 stainless steel, the 
applicant determined the Service Level D limit load collapse pressure 
is 144.1 psi. This pressure is significantly higher than the DSC at-
temperature internal pressure of 45.9 psi and the ASME Code limit-load 
collapse pressure acceptance criteria of 90.2 psi. Correspondingly, 
using the site-specific 237 [deg]F at-temperature material yield 
strength of 24.0 ksi, together with the off-normal at-temperature 
internal pressure of 10.9 psi as a boundary condition, the applicant 
determined the collapse side-drop g-load to be 204 g. This site-
specific collapse side-drop is also much greater than the ASME Code 
limit-load collapse side-drop g-load acceptance criteria of 104 g 
associated with the design basis 500 [deg]F at-temperature material 
yield strength of 19.4 ksi.
    To determine the strain margins of safety for the site-specific 
temperature and pressure, the applicant performed elastic-plastic 
analyses for DSCs 11-15 with the maximized flaws in the OTCP- and ITOP-
to-shell welds. Using the analysis approach in Calculation 11042-0207, 
the applicant calculated the peak equivalent plastic strains of 4.4 
percent and 9.8 percent for the Service Level D internal pressure of 
45.9 psi and the design basis side-drop of 75 g, respectively. For the 
same weld elongation limit of 28 percent, the corresponding strain 
margins of safety are calculated to be 5.36 {(0.28/

[[Page 47200]]

0.044)-1 = 5.36{time}  and 1.86 {(0.28/0.098)-1 = 1.86{time} . Similar 
to the analysis used in Calculation 11042-0207 for a supplement 
qualification of the DSC 16 closure welds with a more conservative 
loading assumption, the applicant also considered 150 percent of the 
site-specific loading to evaluate the weld flaws using a DSC internal 
pressure of 69 psi (45.9 x 1.5 = 69 psi) and side-drop load of 112.5 g. 
The resulting peak equivalent plastic strains are 7.1 percent and 19.0 
percent, which correspond to the strain margins of safety of 2.94 
{(0.28/0.071)-1 = 2.94{time}  and 0.47 {(0.28/0.19)-1 = 0.47{time} , 
respectively. For the MNGP site-specific evaluation, because the 
margins of safety are all positive, the NRC staff concludes that the 
DSCs 11-15 weld strains have additional margins beyond the design basis 
conditions.
    On the basis of the review above, the NRC staff concludes that the 
limit load and elastic-plastic analysis results showed that the welds 
would undergo localized plastic deformation. The applicant's evaluation 
indicated that no weld material rupture or breach of the DSCs 11-15 
confinement boundary at the closure welds is expected because of the 
adequate margins of safety against the weld elongation limits. For this 
reason, the NRC staff has reasonable assurance to conclude that the 
ITCP and OTCP welds of DSCs 11-15 have adequate structural margins of 
safety for the ASME Code Service Level D design criteria, which bound 
the normal, off-normal, and accident (including natural phenomenon) 
conditions for the subject weld structural integrity evaluation. The 
NRC staff also finds that the retrievability of DSCs 11-15 is ensured 
based on the demonstration of adequate weld strain margins of safety 
discussed above.
    Thermal Review for the Requested Exemption: The applicant stated 
that even though nonconforming examinations exist for the primary 
confinement welds, satisfactory completion of the required helium leak 
test conducted on DSCs 11-15 has demonstrated the integrity of the 
primary confinement boundary (ITCP and siphon/vent cover plate) welds. 
These tests specifically demonstrated that the primary confinement 
boundary field welds are ``leak tight'' as defined in ANSI N14.5-1997. 
The applicant stated that, in this respect, the helium leak test 
demonstrated the basic integrity of the primary confinement boundary 
and the lack of a through-weld flaw in the field closure welds that 
would lead to a loss of cavity helium in DSCs 11-15. The applicant 
stated that the field closure welds indirectly support the thermal 
design function by virtue of their confinement function (as 
demonstrated by the helium leak test conducted on DSCs 11-15) which 
assures the helium atmosphere in the DSCs 11-15 cavity is maintained in 
order to support heat transfer. The applicant also stated that the 
satisfactory completion of two required vacuum pump-downs conducted on 
the DSCs demonstrated weld integrity of the ITCP confinement boundary. 
These pump-downs establish a differential pressure across the ITCP and 
siphon/vent block welds of approximately one atmosphere, which exceeds 
the magnitude of the 10 psig design pressure used in stress analyses 
for normal conditions. Although the vacuum pump-down imparts a pressure 
differential in a reverse direction from the confinement function, 
according to the applicant, the pump-down demonstrates the basic 
function of the confinement boundary and the lack of a through-weld 
flaw in the ITCP and siphon/vent block welds sufficient to cause a loss 
of cavity helium when in service.
    The NRC staff reviewed the applicant's exemption request and also 
evaluated its effect on DSCs 11-15 thermal performance. The NRC staff 
concludes that the cask thermal performance is not affected by the 
exemption request because the applicant has shown that a satisfactory 
helium leak test was conducted on DSCs 11-15, which is integral to 
ensuring integrity of the primary confinement boundary. Integrity of 
the primary confinement boundary assures the spent fuel is stored in a 
safe inert environment with unaffected heat transfer characteristics 
that assure peak cladding temperatures remain below allowable limits. 
The NRC staff also concludes that the applicant demonstrated the lack 
of a through-weld flaw in the ITCP and siphon/vent block weld 
sufficient to cause a loss of cavity helium. This satisfies 10 CFR 
72.236(f) which requires that the cask be designed to have adequate 
heat removal capacity without active cooling systems and 10 CFR 
72.122(h) which states that the fuel cladding during storage must be 
protected against degradation and gross rupture. Therefore, based on 
the NRC staff's review of the applicant's evaluation and technical 
justification, the NRC staff finds the exemption request acceptable by 
virtue of the demonstrable structural integrity of the ITCP and siphon/
vent plate welds.
    The NRC staff finds that the thermal function of DSCs 11-15, loaded 
under CoC No. 1004, Amendment No. 10, addressed in the exemption 
request remains in compliance with 10 CFR part 72.
    Shielding and Criticality Safety Review for the Requested 
Exemption: The NRC staff reviewed the criticality safety and radiation 
protection effectiveness of DSCs 11-15 presented in the applicant's 
exemption request. The NRC staff finds that the criticality safety and 
radiation protection of DSCs 11-15 are not affected by the 
nonconforming PT examinations for the following reasons: (1) The 
interior of DSCs 11-15 will continue to prevent water in-leakage which 
means that the system will remain subcritical under all conditions; and 
(2) the nonconforming PT examinations do not affect the radiation 
source term of the spent fuel contents, or the configuration and 
effectiveness of the shielding components of the Standardized 
NUHOMS[supreg] system containing the 61BTH DSC, meaning that the 
radiation protection performance of the system is not altered.
    The NRC staff finds that the criticality safety and shielding 
function of DSCs 11-15, loaded under CoC No. 1004, Amendment No. 10, 
addressed in the exemption request remains in compliance with 10 CFR 
part 72.
    Confinement Review for the Requested Exemption: The objective of 
the confinement evaluation was to confirm that DSCs 11 through 15 
loaded at the MNGP met the confinement-related requirements described 
in 10 CFR part 72. NRC staff relied on the information provided by the 
applicant in their Exemption Request dated October 18, 2017.
    As described in the applicant's ``Exemption Request for 
Nonconforming Dry Shielded Canister Dye Penetrant Examinations'' 
(Exemption Request Enclosure 1), certain elements of the DSCs 11-15 
closure weld PT examinations did not comply with examination procedures 
associated with TS 1.2.5. To support the exemption request, the 
applicant noted that a helium leakage rate test of the closure's 
confinement boundary, including ITCP weld, siphon cover plate weld, and 
vent port cover plate weld, were conducted per TS 1.2.4a and 
demonstrated that the primary confinement barrier field welds met the 
TS acceptance criterion of leaktight as defined by ANSI N14.5-1997. The 
applicant noted that the confinement integrity is not affected by the 
non-compliant PT examination procedures. The NRC staff concludes that 
not performing the PT examination procedures relevant to this exemption 
request would not change the results of the helium leakage test, which 
is integral to ensuring closure confinement

[[Page 47201]]

integrity, and therefore, the closure confinement integrity is 
unaffected. The structural and material acceptability of DSCs 11 
through 15 welds is discussed in the Structural Review and the 
Materials Review described previously.
    It is noted that a dose-related analysis was included as Enclosure 
10 of the Exemption Request. NRC staff neither approves, nor rejects, 
and is not expressing any view related to the material in that 
enclosure, as it did not enter into the evaluation.
    Risk Assessment for the Requested Exemption: In support of the 
applicant's request, the applicant submitted a risk assessment, Jensen 
Hughes Report 016045-RPT-01, ``Risk Assessment of MNGP DSCs 11-15 Welds 
Using NUREG-1864 Methodology'' (Exemption Request Enclosure 11). The 
risk assessment compares the calculated risk of leaving the five DSCs 
in storage ``as is'' at the MNGP ISFSI versus transferring the DSCs 
back into the reactor building to perform PAUT of the welds and then 
returning them to their storage locations. The risk for each potential 
accident, regardless of likelihood, can be generally summarized by the 
following equation:

Initiating Event Frequency (per Year) x Probability of Canister Release 
x Probability of Containment Release x Consequences (Cancer Fatality) = 
Risk

The process to transfer a DSC to the reactor building refueling floor 
for PAUT incurs added potential for accidental drops due to the lifting 
and subsequent lowering operations. For 20-year storage, the risk is 
the sum of all potential accident risks for the duration. Each DSC 
handling operation is independent. For five canisters, the total risk 
value is multiplied by five.
    NUREG-1864, ``A Pilot Probabilistic Risk Assessment of a Dry Cask 
Storage System at a Nuclear Power Plant'' (ADAMS Accession No. 
ML071340012) provides guidance for assessing the risk to the public and 
for identifying the dominant contributors to risk for performing 
probabilistic risk assessments (PRAs) of a dry cask storage system 
located at a nuclear power plant site. NUREG-1864 documents a pilot PRA 
conducted for a dry cask storage system (Holtec International HI-STORM 
100) at a Boiling Water Reactor (BWR) Mark 1 plant. The risk assessment 
estimated the annual off-site risk for one cask in terms of individual 
probability of a prompt fatality and a latent cancer fatality. It does 
not consider risk to workers or future off-site transportation of DSCs.
    The applicant applied the methodology and results in NUREG-1864 to 
perform the risk assessment. The risk assessment compared the 
NUHOMS[supreg] and HI-STORM-100 dry spent fuel storage systems and 
determined the designs are similar with a few basic differences. Both 
storage systems include canisters for confining dry spent fuel. The 
canisters have similar design and dimensions and are made of stainless 
steel of similar thickness and are required to meet the same ASME class 
(ASME B&PV, Section III, and Subsection NB). The HI-STORM 100 system 
consists of a multipurpose canister (MPC) that confines spent fuel 
assemblies, a transfer overpack that provides shielding during canister 
preparation, and a vertical, cylindrical storage overpack that provides 
shielding during long-term storage.
    Both MNGP and Hatch (the plant selected for the Pilot PRA) are BWR, 
Mark 1 plants; therefore, the storage systems are exposed to similar 
handling hazards. The potential drop heights for loaded TCs moving 
across the refueling floor, or lowering from the height of refueling 
floor to the ground floor of the equipment hatch are very similar. The 
potential impact surfaces are also similar.
    The NUHOMS[supreg] system is comprised of a DSC, a TC, and an HSM. 
A transfer trailer is used to move the loaded TC. Two key differences 
exist between the NUHOMS[supreg] and the HI-STORM dry spent fuel 
storage operations. First, the NUHOMS[supreg] TC is placed horizontally 
on the transfer trailer and is not subject to accidental drops when 
moving between the ISFSI and fuel building. Second, transferring 
NUHOMS[supreg] DSC between the TC and the HSM is done horizontally; 
thus, the NUHOMS[supreg] DSC is not subject to any potential vertical 
drop. During storage on an ISFSI pad, the horizontal-storage design of 
the HSM eliminates the risk of tip over caused by seismic activities or 
wind-driven missiles. Aircraft impact on the HSM is limited to only 
large aircrafts and the methodology considered the distance to local 
airfields and planes that operate in the area. The NUREG-1864 frequency 
estimate for meteorite strikes per unit area is used in this 
assessment, and the analysis is adjusted for the larger horizontal 
surface area of the HSM.
    In the risk assessment, the potential radiological consequences are 
based on a comparison of the spent fuel in the MNGP DSC and the spent 
fuel modeled in NUREG-1864. In NUREG-1864, the HI-STORM 100 MPC 
contained 68 BWR fuel assemblies with 10-year-old high-burnup (50 GWD/
MTU) fuel. The MNGP NUHOMS[supreg] DSC contains 61 BWR fuel assemblies 
with 15.5-year-old fuel of 41 GWD/MTU (not high burnup) fuel. The plume 
heat content for a cask release is estimated to be that of the spent 
fuel. NUREG-1864 estimates the maximum decay heat load to be 264 watts 
per assembly. The estimated maximum decay heat load for MNGP DSC is 
approximately 220 watts per assembly. The risk assessment analysis 
assumes that the source term from NUREG-1864 adequately represents or 
bounds those of the MNGP configuration. The NRC staff agrees that this 
is reasonable based on the applicant's assessment which shows NUREG-
1864 radionuclide inventory is 7.0 times higher than that of MNGP DSC.
    The NUREG-1864 evaluation of misload concluded MPC integrity would 
not be affected unless a gross series of errors occurred. The errors 
would have to result in nearly every fuel assembly loaded into the MPC 
being incorrect and insufficiently cooled. NUREG-1864 concluded this 
gross misload scenario was not credible. Therefore, the risk assessment 
did not explore risk from misloading of spent fuel.
    The applicant's risk assessment assumes the annual risk for a DSC 
while stored on the ISFSI would be the same for both alternatives. The 
risk assessment identified three types of mechanical failure that could 
cause significant radiological releases to the environment: drop 
accidents, meteorite strikes, and overflight aircraft accidents. The 
primary difference in risk between the two alternatives, continued 
storage at the ISFSI versus moving a DSC back to the spent fuel pool 
area for PAUT, are potential drop accidents during lifting and lowering 
of a DSC between the ground floor and the height of the refueling 
floor.
    The applicant's risk assessment accounted for possible added risk 
from a potential flaw around the canister lid by assuming the 
probability of lid failure would be same as for the DSC shell in drop 
accidents. This assumption doubles the estimated probability for a 
release from drop accidents. Strain analysis in NUREG-1864 reports the 
most highly stressed regions of the MPC for a drop accident are in 
areas near the base of the cylindrical shell and in the weld joining 
the shell to the baseplate. Since the top side of a canister is not 
expected to experience significant strain, the NRC staff agrees that 
the assumption is conservative and bounds the probability of a release 
occurring following a drop accident.
    The NRC staff reviewed the applicant's risk assessment and agrees

[[Page 47202]]

the mechanical failures identified and the radiological inventory from 
NUREG-1864 would be bounding for each of the MNGP DSCs. The risk 
assessment concludes that the risks are significantly lower than the 
level considered ``negligible'' by the Quantitative Health Guidelines 
(QHG) established in ``Risk-Informed Decisionmaking for Nuclear 
Material and Waste Applications,'' Revision 1 (ADAMS Accession No. 
ML080720238). The QHG considers public individual risk of latent cancer 
fatality risk of less than 2 x 10-6 per year as negligible. 
The pilot PRA (NUREG-1864) concluded that there is no prompt fatality 
risk, and the calculated risk is extremely small. NUREG-1864 reports 
the increase in risk (individual probability of latent cancer fatality) 
from the first year as 1.8 x 10-12, and for subsequent years 
as 3.2 x 10-14 per year per MPC. The total risk for 
Monticello as calculated by Jensen Hughes took into account the 
characteristics of the spent fuel and the site, as well as the 
differences between the MNGP and Hatch ISFSIs. For the five DSCs over a 
period of 20-year storage, risk would be: Alternative 1, continue 
storage as-is, Risk = 1.4 x 10-12; Alternative 2, move DSCs 
back up to the refueling floor for PAUT then return to storage 
location, Risk = 2.3 x 10-12; with a difference in risk 
between the two proposed alternatives of 9.3 x 10-13.
    The assessment of difference in risk between the proposed 
alternatives was performed based on evaluation data from NUREG-1864. 
The MNGP off-site consequence is based on individual risk and not 
absolute population difference. Based on the considerations taken into 
account for the difference between the NUREG-1864 MPC and the MNGP DSCs 
in this assessment, the NRC staff finds the risk assessment calculation 
to be reasonable because the applicant used accepted methods and the 
site-specific considerations were addressed in an appropriately 
conservative manner.
    The purpose of this assessment is to compare the risk associated 
with leaving these DSCs as-is at the ISFSI versus transferring the five 
DSCs back to the refueling floor for PAUT, and then returning them to 
the ISFSI for storage. The process of returning the five DSCs to the 
refueling floor for PAUT incurs additional crane operation. The 
inadvertent drop frequency for heavy loads (NUREG-1774, ``A Survey of 
Crane Operating Experience at U.S. Nuclear Power Plants from 1968 
through 2002'', ADAMS Accession No. ML032060160) is 
5.6x10-5/lift. The probability of release from a DSC drop 
accident, assuming defective weld, is 4.0 x 10-2. This 
operation occurs inside a closed building with probability of release 
value of 1.5 x 10-4. The consequence value for a release is 
3.6 x 10-4. The risk for a drop while lifting a DSC up to 
the refueling floor can be calculated as:

(5.6 x 10-5)(4.0 x 10-2)(1.5 x 
10-4)(3.6 x 10-4) = 1.2 x 10-13 cancer 
fatality/year

The risk for a drop while lowering a DSC (assuming no weld flaw, 
probability of release is 2.0 x 10-2) through the equipment 
hatch back to ground level can be calculated as:

(5.6 x 10-5)(2.0 x 10-2)(1.5 x 
10-4)(3.6 x 10-4) = 6.0 x 10-14 cancer 
fatality/year

The additional risk from performing PAUT for five DSCs would be five 
times the sum of risk for lifting and lowering one DSC.

5 x [(1.2 x 10-13) + (6.0 x 10-14)] = 9.3 x 
10-13 cancer fatality/year

Probabilistic risk assessments are typically used to evaluate risks 
greater than 1.0 x 10-6. In light of the calculated risk 
values, the NRC staff finds the off-site risk as too small to be 
accurately discernable. Based on the discussion presented above, the 
NRC staff concludes that risk to the public for the two options 
provided by Jensen Hughes, ``continued storage as-is'' and ``transfer, 
perform PAUT, and return to storage,'' are essentially equivalent.

Otherwise in the Public Interest

    In considering whether granting the exemption is in the public 
interest, the NRC staff considered the alternative of not granting the 
exemption. If the exemption were not granted, in order to comply with 
the CoC, either (1) DSCs 11-15 would have to be removed from their 
respective HSMs, opened and unloaded, and the contents loaded in new 
DSCs, with each of those new DSCs welded and tested, or (2) removed 
from the HSMs to allow access to the OTCP to be machined off, and the 
ITCP weld machined down to the root weld; and each DSC, ITCP and OTCP 
inspected to determine if there was any damage as a result of the 
machining (which would then necessitate the actions detailed in option 
1); or (3) conduct PAUT by opening the HSMs to conduct in-situ testing 
(which is limited to less than 360[deg] of the weld circumference) or 
transferring to a TC for testing on the ISFSI pad or in the reactor 
building (essentially Alternative 2 in the Risk Assessment). Options 1 
and 2 would entail a higher risk of cask handling accidents, additional 
personnel exposure, and greater cost to the applicant. As noted above 
in the Risk Assessment, Option 3 does not increase the risk by a 
discernible amount. All options would generate additional radioactive 
contaminated material and waste from operations. For options 1 and 2, 
the lid would have to be removed, which would generate cuttings from 
removing the weld material that could require disposal as contaminated 
material. For option 3, radioactive wastes would be generated from 
radioactively contaminated consumables and anti-contamination clothing 
used during the examination. Also, radioactive waste would be generated 
from the cleanup of any coupling fluid (of the PAUT) that it combines 
with and then transports resulting in contamination from the surface of 
the DSC. 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.
    The proposed exemption to permit continued storage of DSCs 11-15 in 
their respective HSMs for the service life of the canisters at the MNGP 
ISFSI is consistent with NRC's mission to protect public health and 
safety. Approving the requested exemption reduces the opportunity for a 
release of radioactive material compared to the alternatives to the 
proposed action, because there will be no operations involving the 
opening of the DSCs, which confine the spent nuclear fuel, and there 
will be no operations involving the opening of the HSMs potentially 
exposing radioactive waste to the environment. Therefore, the exemption 
is 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 exemption meets the categorical exclusion in 10 CFR 
51.22(c)(25).
    Granting this exemption from 10 CFR 72.212(a)(2), 72.212(b)(3), 
72.212(b)(5)(i), 72.214, and 72.212(b)(11) only relieves the applicant 
from the inspection or surveillance requirements associated with 
performing PT examinations with regard to meeting TS 1.2.5 of 
Attachment A of CoC No. 1004. A categorical exclusion for inspection or 
surveillance requirements is provided under 10 CFR 51.22(c)(25)(vi)(C) 
if the criteria in 10

[[Page 47203]]

CFR 51.22(c)(25)(i)-(v) are also satisfied. In its review of the 
exemption request, the NRC staff determined, as discussed above, that, 
under 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 or consequences from radiological accidents such 
as a gross leak from the closure welds, because the exemption neither 
reduces the ability of the closure welds to confine radioactive 
material nor creates new accident precursors at the MNGP ISFSI. 
Accordingly, this 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 following methods, as 
indicated.

------------------------------------------------------------------------
                 Document                       ADAMS  accession No.
------------------------------------------------------------------------
Federal Register Notice Issuing Exemption  ML16159A227
 from Nonconforming Dye Penetrant
 Examinations of Dry Shielded Canister
 (DSC) 16, June 8, 2016.
Exemption Request for Nonconforming Dye    ML17296A205
 Penetrant Examinations of Dry Shielded
 Canisters (DSCs) 11 through 15, October
 18, 2017.
First Request for Additional Information   ML18065A545
 for Review of Exemption Request for Five
 Nonconforming Dry Shielded Canisters 11
 through 15 (CAC No. 001028, Docket No.
 72-58, EPID L-2017-LLE-0029), March 6,
 2018.
Monticello Nuclear Generating Plant--      ML18100A173
 Response to Request for Additional
 Information Regarding Exemption Request
 for Nonconforming Dye Penetrant
 Examinations of Dry Shielded Canisters
 (DSCs) 11 through 15, April 5, 2018.
Supplement to Exemption Request for        ML18151A870
 Nonconforming Dye Penetrant Examinations
 of Dry Shielded Canisters (DSCs) 11
 through 15 (CAC No. 001028, EPID L-2017-
 LLE-0029).
NUREG-1774, ``A Survey of Crane Operating  ML032060160
 Experience at U.S. Nuclear Power Plants
 from 1968 through 2002''.
Risk-Informed Decisionmaking for Nuclear   ML080720238
 Material and Waste Applications,
 Revision 1.
NUREG-1536, Revision 1 ``Standard Review   ML101040620
 Plan for Spent Fuel Dry Storage Systems
 at a General License Facility''.
NUREG-1864, ``A Pilot Probabilistic Risk   ML071340012
 Assessment of a Dry Cask Storage System
 at a Nuclear Power Plant''.
Attachment A, Technical Specifications,    ML17338A114
 Transnuclear, Inc., Standardized
 NUHOMS[supreg] Horizontal Modular
 Storage System Certificate of Compliance
 No. 1004, Renewed Amendment No. 10,
 Revision 1.
------------------------------------------------------------------------

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 
applicant an exemption from the requirements of 10 CFR 72.212(a)(2), 
72.212(b)(3), 72.212(b)(5)(i), 72.212(b)(11), and 72.214 only with 
regard to meeting TS 1.2.5 of Attachment A of CoC No. 1004 for DSCs 11-
15.
    This exemption is effective upon issuance.

    Dated at Rockville, Maryland, this 13th day September 2018.

    For the Nuclear Regulatory Commission.
John McKirgan,
Branch Chief, Spent Fuel Licensing Branch, Division of Spent Fuel 
Management, Office of Nuclear Material Safety and Safeguards.
[FR Doc. 2018-20283 Filed 9-17-18; 8:45 am]
BILLING CODE 7590-01-P



                                               47192                     Federal Register / Vol. 83, No. 181 / Tuesday, September 18, 2018 / Notices

                                               NATIONAL SCIENCE FOUNDATION                             exemption in response to a request                    Material Safety and Safeguards, U.S.
                                                                                                       submitted by Xcel Energy on October                   Nuclear Regulatory Commission,
                                               Advisory Committee for International                    18, 2017, from meeting Technical                      Washington, DC 20555–0001; telephone:
                                               Science and Engineering; Notice of                      Specification (TS) 1.2.5 of Attachment A              301–415–6825; email: Christian.Jacobs@
                                               Meeting                                                 of Certificate of Compliance (CoC) No.                nrc.gov.
                                                 In accordance with the Federal                        1004, Amendment No. 10, which                         SUPPLEMENTARY INFORMATION:
                                               Advisory Committee Act (Pub. L. 92–                     requires that all dry shielded canister
                                                                                                       (DSC) closure welds, except those                     I. Background
                                               463, as amended), the National Science
                                                                                                       subjected to full volumetric inspection,                 Northern States Power Company-
                                               Foundation (NSF) announces the
                                                                                                       be dye penetrant tested in accordance                 Minnesota, doing business as Xcel
                                               following meeting:
                                                 Name and Committee Code: Advisory                     with the requirements of American                     Energy (Xcel Energy, or the applicant) is
                                               Committee for International Science and                 Society of Mechanical Engineers                       the holder of Renewed Facility
                                               Engineering Meeting (AC–ISE) (#25104).                  (ASME) Boiler and Pressure Vessel                     Operating License No. DPR–22, which
                                                 Date and Time: Monday, October 29,                    (B&PV) Code Section III, Division 1,                  authorizes operation of the MNGP, Unit
                                               2018; 9:00 a.m. to 4:45 p.m. (EDT),                     Article NB–5000. This exemption                       No. 1, in Wright County, Minnesota,
                                               Tuesday, October 30, 2018; 9:00 a.m. to                 applies to five loaded Standardized                   pursuant to part 50 of title 10 of the
                                               1:00 p.m. (EDT).                                        NUHOMS® 61BTH, Dry Shielded                           Code of Federal Regulations (10 CFR),
                                                 Place: National Science Foundation,                   Canisters (DSCs) 11 through 15, at the                ‘‘Domestic Licensing of Production and
                                               2415 Eisenhower Avenue, Alexandria,                     Monticello Nuclear Generating Plant                   Utilization Facilities.’’ The license
                                               VA 22314.                                               (MNGP) Independent Spent Fuel                         provides, among other things, that the
                                                 To help facilitate your entry into the                Storage Installation (ISFSI).                         facility is subject to all rules,
                                               NSF building, please contact Victoria                   ADDRESSES: Please refer to Docket ID
                                                                                                                                                             regulations, and orders of the NRC now
                                               Fung (vfung@nsf.gov) on or prior to                     NRC–2018–0207 when contacting the                     or hereafter in effect.
                                               October 24, 2018.                                                                                                Consistent with 10 CFR part 72,
                                                                                                       NRC about the availability of
                                                 Type of Meeting: Open.                                                                                      subpart K, ‘‘General License for Storage
                                                                                                       information regarding this document.
                                                 Contact Person: Simona Gilbert, AC–                                                                         of Spent Fuel at Power Reactor Sites,’’
                                                                                                       You may obtain publicly-available
                                               ISE Executive Secretary and Staff                                                                             a general license is issued for the storage
                                                                                                       information related to this document
                                               Associate for Budget, National Science                                                                        of spent fuel in an ISFSI at power
                                                                                                       using any of the following methods:
                                               Foundation, 2415 Eisenhower Avenue,                                                                           reactor sites to persons authorized to
                                                                                                          • Federal Rulemaking Website: Go to
                                               Alexandria, Virginia, 22314; Telephone:                                                                       possess or operate nuclear power
                                                                                                       http://www.regulations.gov and search
                                               703–292–8710.                                                                                                 reactors under 10 CFR part 50. The
                                                                                                       for Docket ID NRC–2018–0207. Address
                                                 Purpose of Meeting: To provide                                                                              applicant is authorized to operate a
                                                                                                       questions about Docket IDs in
                                               advice, recommendations and counsel                                                                           nuclear power reactor under 10 CFR
                                                                                                       Regulations.gov to Jennifer Borges;                   part 50, and holds a 10 CFR part 72
                                               on major goals and policies pertaining                  telephone: 301–287–9127; email:
                                               to international programs and activities.                                                                     general license for storage of spent fuel
                                                                                                       Jennifer.Borges@nrc.gov. For technical                at the MNGP ISFSI. Under the terms of
                                               Agenda                                                  questions, contact the individual(s)                  the general license, the applicant stores
                                               • Updates on OISE activities                            listed in the FOR FURTHER INFORMATION                 spent fuel at its ISFSI using the TN
                                               • Discussion on International Strategic                 CONTACT section of this document.
                                                                                                                                                             Americas LLC Standardized NUHOMS®
                                                 Plan Working Group                                       • NRC’s Agencywide Documents                       dry cask storage system in accordance
                                               • Updates on MULTIplying Impact                         Access and Management System                          with CoC No. 1004, Amendments No. 9
                                                 Leveraging International Expertise in                 (ADAMS): You may obtain publicly-                     and No. 10. As part of the dry storage
                                                 Research (MULTIPLIER)                                 available documents online in the                     system, the DSC (of which the closure
                                               • Updates on IRES Evaluation                            ADAMS Public Documents collection at                  welds are an integral part) ensures that
                                               • Discussion on International Strategic                 http://www.nrc.gov/reading-rm/                        the dry storage system can meet the
                                                 Plan                                                  adams.html. To begin the search, select               functions of criticality safety,
                                               • Meet with NSF leadership                              ‘‘ADAMS Public Documents’’ and then                   confinement boundary, shielding,
                                                 Dated: September 12, 2018.                            select ‘‘Begin Web-based ADAMS                        structural support, and heat transfer.
                                               Crystal Robinson
                                                                                                       Search.’’ For problems with ADAMS,
                                                                                                       please contact the NRC’s Public                       II. Request/Action
                                               Committee Management Officer.
                                                                                                       Document Room (PDR) reference staff at                   The applicant has requested an
                                               [FR Doc. 2018–20170 Filed 9–17–18; 8:45 am]
                                                                                                       1–800–397–4209, 301–415–4737, or by                   exemption from the requirements of 10
                                               BILLING CODE 7555–01–P
                                                                                                       email to pdr.resource@nrc.gov. The                    CFR 72.212(a)(2), 10 CFR 72.212(b)(3),
                                                                                                       ADAMS accession number for each                       10 CFR 72.212(b)(5)(i), 10 CFR
                                                                                                       document referenced (if it is available in            72.212(b)(11), and 10 CFR 72.214 that
                                               NUCLEAR REGULATORY                                      ADAMS) is provided the first time that                require compliance with the terms,
                                               COMMISSION                                              it is mentioned in this document. In                  conditions, and specifications of CoC
                                               [Docket Nos. 72–58 and 50–263; NRC–2018–                addition, for the convenience of the                  No. 1004, Amendment No. 10, for the
                                               0207]                                                   reader, the ADAMS accession numbers                   Standardized NUHOMS® Horizontal
                                                                                                       are provided in a table in the                        Modular Storage System, to allow
                                               Xcel Energy, Monticello Nuclear                         ‘‘Availability of Documents’’ section of              continued storage of DSCs 11–15 in
                                               Generating Plant; Independent Spent                     this document.                                        their respective Horizontal Storage
daltland on DSKBBV9HB2PROD with NOTICES




                                               Fuel Storage Installation                                  • NRC’s PDR: You may examine and                   Modules (HSMs). This would permit the
                                               AGENCY:  Nuclear Regulatory                             purchase copies of public documents at                continued storage of those five DSCs for
                                               Commission.                                             the NRC’s PDR, Room O1–F21, One                       the service life of the canisters.
                                                                                                       White Flint North, 11555 Rockville                    Specifically, the exemption would
                                               ACTION: Exemption; issuance.
                                                                                                       Pike, Rockville, Maryland 20852.                      relieve the applicant from meeting TS
                                               SUMMARY:The U.S. Nuclear Regulatory                     FOR FURTHER INFORMATION CONTACT:                      1.2.5 of Attachment A of CoC No. 1004
                                               Commission (NRC) is issuing an                          Christian Jacobs, Office of Nuclear                   (ADAMS Accession No. ML17338A114),


                                          VerDate Sep<11>2014   19:14 Sep 17, 2018   Jkt 244001   PO 00000   Frm 00069   Fmt 4703   Sfmt 4703   E:\FR\FM\18SEN1.SGM   18SEN1


                                                                         Federal Register / Vol. 83, No. 181 / Tuesday, September 18, 2018 / Notices                                            47193

                                               which requires that all DSC closure                     and before applying the changes                       DSCs 11–15 in their respective HSMs at
                                               welds, except those subjected to full                   authorized by an amended CoC to a cask                the MNGP ISFSI for the service life of
                                               volumetric inspection, be dye penetrant                 loaded under the initial CoC or an                    the canisters. As detailed below, NRC
                                               tested in accordance with the                           earlier amended CoC, which establish                  staff reviewed the exemption request to
                                               requirements of the ASME B&PV Code                      that the cask, once loaded with spent                 determine whether granting of the
                                               Section III, Division 1, Article NB–5000.               fuel or once the changes authorized by                exemption would cause potential for
                                               Technical Specification 1.2.5 further                   an amended CoC have been applied,                     danger to life, property, or common
                                               requires that the dye penetrant test (PT)               will conform to the terms, conditions,                defense and security.
                                               acceptance standards be those described                 and specifications of a CoC or an                     Review of the Requested Exemption
                                               in Subsection NB–5350 of the ASME                       amended CoC listed in 10 CFR 72.214;
                                               BP&V Code.                                                • 10 CFR 72.212(b)(11), which states,                  The NUHOMS® system provides
                                                  Xcel Energy loaded spent nuclear fuel                in part, that the licensee shall comply               horizontal dry storage of canisterized
                                               into six 61BTH DSCs starting in                         with the terms, conditions, and                       spent fuel assemblies in an HSM. The
                                               September 2013. Subsequent to the                       specifications of the CoC and, for those              cask storage system components for
                                               loading, it was discovered that certain                 casks to which the licensee has applied               NUHOMS® consist of a reinforced
                                               elements of the PT examinations, which                  the changes of an amended CoC, the                    concrete HSM and a DSC vessel with an
                                               were performed on the DSCs to verify                    terms, conditions, and specifications of              internal basket assembly that holds the
                                               the acceptability of the closure welds,                 the amended CoC; and                                  spent fuel assemblies. The HSM is a
                                               do not comply with the requirements of                    • 10 CFR 72.214, which lists the                    low-profile, reinforced concrete
                                               TS 1.2.5. All six DSCs were affected.                   approved spent fuel storage casks.                    structure designed to withstand all
                                               Five of the six DSCs (numbers 11–15)                                                                          normal condition loads, as well as
                                               had already been loaded in the HSMs                     III. Discussion                                       abnormal condition loads created by
                                               when the discrepancies were                                Pursuant to 10 CFR 72.7, the                       natural phenomena such as earthquakes
                                               discovered. DSC 16 remained on the                      Commission may, upon application by                   and tornadoes. It is also designed to
                                               reactor building refueling floor in a                   any interested person or upon its own                 withstand design basis accident
                                               transfer cask (TC). On June 8, 2016, NRC                initiative, grant such exemptions from                conditions. The Standardized
                                               granted an exemption (ADAMS                             the requirements of the regulations of 10             NUHOMS® Horizontal Modular Storage
                                               Accession No. ML16159A227) from 10                      CFR part 72 as it determines are                      System has been approved for storage of
                                               CFR 72.212(a)(2), 10 CFR 72.212(b)(3),                  authorized by law and will not endanger               spent fuel under the conditions of CoC
                                               10 CFR 72.212(b)(5)(i), 10 CFR                          life or property or the common defense                No. 1004. The DSCs under
                                               72.212(b)(11), and 10 CFR 72.214 for                    and security and are otherwise in the                 consideration for exemption were
                                               DSC 16 only with regard to meeting TS                   public interest.                                      loaded under CoC No. 1004,
                                               1.2.5 of Attachment A of CoC No.1004,                                                                         Amendment No. 10.
                                                                                                       Authorized by Law
                                               Amendment No. 10. The exemption                                                                                  The NRC has previously approved the
                                               granted on June 8, 2016, restored DSC                      This exemption would permit the                    Standardized NUHOMS® Horizontal
                                               16 to compliance with 10 CFR part 72                    continued storage of DSCs 11–15 at the                Modular Storage System. The requested
                                               and allowed Northern States Power                       MNGP ISFSI for the service life of the                exemption does not change the
                                               Company-Minnesota to transfer DSC 16                    canisters by relieving the applicant of               fundamental design, components,
                                               into an HSM for continued storage at                    the requirement to meet the PT                        contents, or safety features of the storage
                                               MNGP ISFSI for the service life of the                  requirements of TS 1.2.5 of Attachment                system. The NRC staff has evaluated the
                                               canister.                                               A of CoC No. 1004. The provisions in                  applicable potential safety impacts of
                                                  In a letter dated October 18, 2017                   10 CFR part 72 from which the                         granting the exemption to assess the
                                               (ADAMS Accession No. ML17296A205)                       applicant is requesting exemption                     potential for danger to life or property
                                               (Exemption Request), as supplemented                    require the licensee to comply with the               or the common defense and security; the
                                               in responses to NRC requests for                        terms, conditions, and specifications of              evaluation and resulting conclusions are
                                               additional information dated April 5,                   the CoC for the approved cask model it                presented below. The potential impacts
                                               2018 (ADAMS Accession No.                               uses. Section 72.7 allows the NRC to                  identified for this exemption request
                                               ML18100A173) (RAI Response 1) and                       grant exemptions from the requirements                were in the areas of materials, structural
                                               May 31, 2018 (ADAMS Accession No.                       of 10 CFR part 72. As explained below,                integrity, thermal, shielding, criticality,
                                               ML18151A870) (RAI Response 2), the                      the proposed exemption will not                       and confinement capability.
                                               applicant requested an exemption from                   endanger life or property, or the                        Materials Review for the Requested
                                               the following requirements to allow                     common defense and security, and is                   Exemption: The applicant asserted that
                                               continued storage of the remaining                      otherwise in the public interest.                     there is a reasonable assurance of safety
                                               DSCs 11–15 in their respective HSMs at                  Issuance of this exemption is consistent              to grant the requested exemption to
                                               the MNGP ISFSI:                                         with the Atomic Energy Act of 1954, as                continue the storage of DSCs 11–15 in
                                                  • 10 CFR 72.212(a)(2), which states                  amended, and not otherwise                            their respective HSMs. The applicant’s
                                               that this general license is limited to                 inconsistent with NRC’s regulations or                assertion of reasonable assurance of
                                               storage of spent fuel in casks approved                 other applicable laws. Therefore, the                 safety is based on the following factors:
                                               under the provisions of part 72;                        exemption is authorized by law.                          • Reasonable assurance of weld
                                                  • 10 CFR 72.212(b)(3), which states                                                                        integrity;
                                                                                                       Will Not Endanger Life or Property or
                                               that the general licensee must ensure                                                                            • Low dose consequences for a DSC
                                               that each cask used by the general                      the Common Defense and Security
                                                                                                                                                             in storage; and
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                                               licensee conforms to the terms,                           This exemption would relieve the                       • Low risk to the public.
                                               conditions, and specifications of a CoC                 applicant from meeting TS 1.2.5 of                       The applicant further stated that there
                                               or an amended CoC listed in 10 CFR                      Attachment A of CoC No. 1004, which                   is reasonable assurance of weld integrity
                                               72.214;                                                 requires PT examinations to be                        based on the existing Quality Assurance
                                                  • 10 CFR 72.212(b)(5)(i), which                      performed on the DSCs to verify the                   (QA) documentation, engineering
                                               requires that the general licensee                      acceptability of the closure welds, and               analysis, and expert evaluations, which
                                               perform written evaluations, before use                 would permit the continued storage of                 demonstrate that the subject DSC welds


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                                               47194                     Federal Register / Vol. 83, No. 181 / Tuesday, September 18, 2018 / Notices

                                               possess sufficient quality to perform                   indicate that all of the fuel assemblies              and closure welds because the extent of
                                               their design functions due to the                       loaded into DSCs 11–15 met the TS                     fatigue cycling experienced by the
                                               following:                                              requirements (TS Table 1–1t) for                      canister is below the threshold which
                                                  • Fuel cladding integrity is                         cladding integrity and no damaged fuel                the ASME B&PV Code Section III has
                                               maintained, as no damaged fuel was                      was loaded. The applicant stated that                 established.
                                               loaded and no unexpected dose                           the integrity of the fuel was further                    The NRC staff have previously
                                               readings were observed during drying                    demonstrated by the fact that no                      reviewed the design of the NUHOMS®
                                               operations.                                             unexpected dose rate readings were                    61BTH DSC included in the UFSAR.
                                                  • The weld design assures that there                 observed during the vacuum drying                     The NRC staff verified that the top cover
                                               are no pinhole leaks and there is no                    processes of DSCs 11–15.                              plate and weld material are stainless
                                               credible process for service-induced                       The NRC staff reviewed the                         steel and the only welds subject to the
                                               flaws.                                                  information provided by the applicant                 outside environment are the outer layer
                                                  • The material, including the DSC                    on the characteristics of the spent fuel              of the OTCP weld and the TPP weld.
                                               shell, lids and weld filler, met quality                loaded in DSCs 11–15. The NRC staff                   The NRC staff verified that the
                                               requirements and quality welds were                     also reviewed the loading records for                 differential pressure across the top cover
                                               ensured by welding process                              the loading campaign and confirmed                    plates is minimal and consequently the
                                               qualification, welder qualification and                 that (1) no damaged fuel assemblies                   reduction in cross section from plastic
                                               the use of an automated welding process                 were loaded in the DSCs; (2) only one                 strain is not credible. The NRC staff
                                               specifically designed for the                           fuel assembly had burnup that                         have reviewed the assessment of fatigue
                                               application.                                            marginally exceeded the 45 GWD/MTU                    and determined that the DSCs are not
                                                  • In-process visual inspections of                   criterion for high burnup fuel however,               subjected to cyclic loading that requires
                                               welds performed by the welders,                         the cladding of the fuel assembly was                 a fatigue analysis. Based on the NRC
                                               Quality Control (QC) visual examination                 shown to be intact through cask loading               staff’s previous analysis of the DSC weld
                                               (VT) inspections of fit-ups and welds,                  reports and supporting radiochemistry                 design, the NRC staff determined that
                                               and the vacuum hold, helium pressure                    reports; and (3) no unexpected dose                   the applicant’s assessment of the weld
                                               and helium leak test all ensured                        readings were observed in the loading                 design is accurate and there is no
                                               confinement and quality of the welds.                   campaign. Based on the review of the                  credible mechanism for the propagation
                                                  • Strain margins for the DSC welds                   information from the loading campaign,                of an existing weld flaw to result in a
                                               were demonstrated by structural                         the NRC staff confirmed that the                      through weld thickness penetration that
                                               analysis assuming flaw distributions                    characteristics of the fuel loaded in the             would result in a leak.
                                               conservatively derived from the Phased                  DSCs included in the exemption request                   Material and Welding Process: The
                                               Array Ultrasonic Testing (PAUT)                         were accurately described.                            applicant stated that procurement
                                               examination of DSC 16.                                     Weld Design: The applicant stated                  records such as certified material test
                                                  • Based on the DSCs 11–15 site-                      that the updated final safety analysis                reports (CMTRs) demonstrate that the
                                               specific heat load conditions, additional               report (UFSAR) only describes weld                    canisters, lids, and weld filler materials
                                               margin exists to account for any                        failure in terms of a possible pinhole                met design standards and quality
                                               remaining flaw uncertainty.                             leak in individual weld layers. The                   requirements, thereby assuring
                                                  The NRC materials review for the                     applicant further stated that the UFSAR               compatibility between materials and
                                               requested exemption focused on the                      assumes or stipulates that pinholes may               satisfactory material performance
                                               applicant’s assertion of reasonable                     exist in individual layers but the                    characteristics (e.g., material strength).
                                               assurance of weld integrity and each of                 UFSAR makes no explicit mention                          The applicant stated that the weld
                                               the supporting assertions of: (1) Fuel                  about how a pinhole leak in a weld                    closures of DSCs 11–15 were performed
                                               cladding integrity; (2) weld design; (3)                layer is formed, whether it occurs                    under a 10 CFR part 50 Appendix B QA
                                               material and welding process; (4) tests                 during the weld formation or by                       program, such that the canister integrity
                                               performed; (5) adequate strain margins                  subsequent canister loading operations,               is assured. The applicant stated that
                                               to accommodate flaws; and (6)                           fatigue cycles during storage, or                     welding materials were procured to
                                               additional strain margins in welds. A                   accidents. The applicant stated that the              quality requirements, welding processes
                                               specific review of each of the supporting               existence of pinhole leaks is a non-                  were developed and qualified for the
                                               statements is provided in the following                 mechanistic assumption of the UFSAR;                  given configuration, and welders were
                                               sections.                                               and there is no underlying malfunction                appropriately qualified to the ASME
                                                  Fuel Cladding Integrity: The applicant               that causes its formation.                            B&PV Code requirements. Finally, the
                                               provided information on the nature of                      The applicant stated that, once in                 applicant stated that welding
                                               the spent nuclear fuel in DSCs 11–15 to                 storage, there is no credible failure                 parameters were specified in associated
                                               demonstrate that the fuel cladding                      mechanism of the DSC top cover plate                  procedures and monitored as required.
                                               fission product barrier is intact and any               closure welds that would adversely                       In addition to the original weld head
                                               postulated canister weld leak would                     affect DSC confinement because (1) the                video review conducted in conjunction
                                               have an insignificant effect on                         top cover plate and weld material are                 with the DSC 16 exemption request, the
                                               radioactive release. At the time of                     stainless steel and the only welds                    applicant included another examination
                                               loading in 2013, the applicant stated                   subject to the outside environment are                of the weld head video and the general
                                               that the combined decay heat load in                    the outer layer of the outer top cover                area videos taken during the 2013 cask
                                               the limiting DSC did not exceed 10.96                   plate (OTCP) weld and the test port plug              loading campaign. Based on the
                                               kilowatts. In addition, only one of the                 (TPP) weld; (2) a reduction in cross                  examination of the videos, the applicant
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                                               305 loaded fuel assemblies was                          section from plastic strain is not                    made a correlation between weld
                                               considered to be high burnup, with a                    applicable to the top cover plate welds               techniques and typical weld flaw
                                               maximum recorded burnup of 45.12                        because the differential pressure across              characteristics such as those identified
                                               gigawatt days per metric ton of uranium                 the top cover plates conditions is                    in the PAUT of the inner top cover plate
                                               (GWD/MTU) (in DSC 15). The applicant                    minimal (less than one atmosphere);                   (ITCP) and OTCP welds from DSC 16.
                                               stated that cask loading reports and                    and (3) the mechanism of cyclic loading               The applicant provided an assessment
                                               supporting radiochemistry records                       is not applicable to the top cover plate              conducted by Structural Integrity


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                                                                         Federal Register / Vol. 83, No. 181 / Tuesday, September 18, 2018 / Notices                                          47195

                                               Associates, Inc. (SIA), which concluded                 multipass lid-to-shell weld of an                     a weld meeting visual examination
                                               that defects would be limited in the                    austenitic stainless steel canister                   criteria was developed; and
                                               through thickness dimension to the                      designed and fabricated in accordance                    • Helium leakage tests to verify the
                                               thickness of a single bead. The applicant               with the ASME B&PV Code Section III                   confinement integrity function and, to
                                               also stated that, even considering the                  Subsection NB (Class 1 components), no                some extent, the structural integrity
                                               possibility that any given layer of weld                flaws of significant size will exist such             function of the DSC welds.
                                               may have a leak through that layer, the                 that the flaws could impair the                          The applicant provided an extent of
                                               licensing basis criterion stated in the                 structural strength or confinement                    condition assessment as Appendix D of
                                               UFSAR Section 3.3.2.1 assures that the                  capability of the weld. For a spent                   Enclosure 1 of the Exemption Request.
                                               chance of pinholes being in alignment                   nuclear fuel canister, such a flaw would              The applicant stated that the extent of
                                               on successive independently-deposited                   be the result of improper fabrication or              condition assessment was focused on:
                                               weld layers is not credible.                            welding technique, as service-induced                    • Compliance with welding
                                                  As stated above, the NRC staff have                  flaws under normal and off-normal                     administrative requirements;
                                               previously reviewed the design of the                   conditions of storage are not credible.                  • Technical specification required
                                               NUHOMS® 61BTH DSC included in the                          The NRC staff notes that per the                   testing of welds; and
                                               UFSAR. The NRC staff reviewed the                       guidance in NUREG–1536, Revision 1,                      • Weld depth measurements for outer
                                               materials used in the construction of                   Section 8.4.7.4, the large structural lid-            top cover plate welds.
                                               DSCs 11–15 and the NRC staff                            to-shell weld designs fabricated from                    The NRC staff reviewed the
                                               confirmed that the materials used met                   austenitic materials may be tested using              information provided in the application
                                               the specifications called out in the                    non-destructive examination methods                   and confirmed that the applicant
                                               NUHOMS® 61BTH DSC design. The                           such as a volumetric ultrasonic test (UT)             provided documentation that the
                                               NRC staff reviewed the CMTRs and                        or a multi-pass PT. If a multiple-pass PT             welding administrative requirements
                                               confirmed that the materials met                        examination is utilized in lieu of UT                 were met, as follows: (1) Welding
                                               specified compositional and mechanical                  inspection, a stress reduction factor of              procedures were available at the job site
                                               property requirements.                                  0.8 for weld strength is imposed. In the              for welding operators to follow; (2) weld
                                                  The NRC staff reviewed, ‘‘TRIVIS Inc.                absence of valid PT examinations of the               surface preparations were completed
                                               Welding Procedure Specification (WPS)                   closure welds for DSCs 11–15, the                     such that the weld surface was dry and
                                               SS–8–M–TN, Revision 10,’’ (Enclosure 2                  applicant asserted that the helium leak               free of oil, grease, weld spatter, rust,
                                               to RAI Response 1) which was used for                   rate tests performed on all DSCs and the              slag, sand, discontinuities, or other
                                               the machine welding of the ITCP and                     PAUT results for DSC 16, which show                   extraneous material; (3) weld crown
                                               the OTCP as well as, ‘‘TRIVIS Inc. WPS                  that weld defects are limited to the                  height for the ITCP and vent/siphon
                                               SS–8–A–TN, Revision 8,’’ (RAI                           height of one weld bead, support the                  port were verified; and (4) welds for the
                                               Response 1 Enclosure 3) used for                        claim that DSCs 11–15 do not have                     ITCP, OTCP and the vent and siphon
                                               manual welding of the ITCP and the                      flaws that would impair the structural                ports were all verified.
                                               OTCP. The NRC staff compared WPS                        strength or confinement capability.                      The NRC staff reviewed the
                                               SS–8–M–TN, Revision 10 and WPS SS–                         The NRC staff reviewed the                         information provided in the application
                                               8–A–TN, Revision 8 to the essential                     information provided by the applicant                 and confirmed that the applicant
                                               variables required for the gas tungsten                 including the DSC lid-to-shell closure                provided documentation for the TS
                                               arc welding (GTAW) in ASME Section                      weld design for the ITCP and the OTCP,                required tests performed on DSCs 11–
                                               IX Part QW Welding, Article II Welding                  the manual and machine GTAW WPSs,                     15. The NRC staff verified that the
                                               Procedure Qualifications, Table QW–                     the helium leak testing results for DSCs              application included documentation
                                               256 and Article IV Welding Data,                        11–15 and the PAUT results for DSC 16.                showing that (1) hydrogen monitoring
                                               Subsection QW–400 Variables. The NRC                    The NRC staff concluded that the design               was properly performed while welding
                                               staff determined that the WPS SS–8–M–                   of the DSC closure weld and the GTAW                  in accordance with TS 1.1.11; (2)
                                               TN, Revision 10 and WPS SS–8–A–TN,                      WPSs used to weld the ITCP and the                    pressure testing of the DSC shell to ITCP
                                               Revision 8 are acceptable because all of                OTCP are unlikely to result in weld                   weld was conducted in accordance with
                                               the essential variables identified in                   flaws that could impair the structural                TS 1.1.12.4; (3) two cycles of vacuum
                                               ASME Section IX for GTAW WPSs were                      strength or confinement capability of                 drying and verification were conducted
                                               included and the range of permissible                   the weld. The NRC staff concluded that                at a vacuum less than 2.8 torr and were
                                               values were specified.                                  the helium leak testing results for DSCs              maintained for times longer than 30
                                                  The NRC staff reviewed, ‘‘TRIVIS, Inc.               11–15 confirmed that there were no                    minutes in accordance with TS 1.2.2; (4)
                                               Procedure Qualification Record (PQR)                    flaws that impaired the confinement                   the DSCs were backfilled with helium
                                               PQR–1, Revision 2’’ (Enclosure 4 to RAI                 capability of the DSC 11–15 ITCP welds.               and to a pressure of 17.2 ± 1.0 psi for
                                               Response 1). The NRC staff compared                     The NRC staff concluded that the PAUT                 a time of at least 30 minutes in
                                               the testing documented in PQR–1,                        results for DSC 16 is sufficient to show              accordance with TS 1.2.3a; and (5)
                                               Revision 2 against ASME Section IX                      that the GTAW of the ITCP and OTCP                    helium backfilling, pressure verification
                                               Part QW Welding, Article I Welding                      welds do not result in defects that                   and leak testing were conducted in
                                               General Requirements. The NRC staff                     would impair structural strength or                   accordance with American National
                                               determined that PQR–1 Revision 2 was                    confinement capability of the DSC                     Standards Institute (ANSI) N14.5–1997
                                               acceptable because all the testing                      closure welds.                                        and leak rates less than 1.0 × 10¥7 ref
                                               necessary to qualify WPS SS–8–M–TN,                        Tests Performed: The applicant stated              cubic centimeters/sec were documented
                                               Revision 10 and WPS SS–8–A–TN,                          that a number of independent tests were               for DSCs 11–15 in accordance with TS
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                                               Revision 8 were performed with                          conducted on the DSC 11–15 welds                      1.2.4a.
                                               satisfactory results and documented in                  which verify that adequate welds were                    The NRC staff confirmed that the
                                               PQR–1, Revision 2.                                      performed on DSCs 11–15. The                          weld depth measurements for the OTCP
                                                  As documented in NUREG–1536,                         applicant stated that these tests include:            were conducted at four locations around
                                               Revision 1, Section 8.9.1 (ADAMS                           • In-process visual examination and                the weld circumference. The NRC staff
                                               Accession No. ML101040620) the NRC                      QC visual examinations to demonstrate                 confirmed that the weld depth
                                               previously determined that for a                        that weld processes were followed and                 (dimension of the weld throat)


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                                               47196                     Federal Register / Vol. 83, No. 181 / Tuesday, September 18, 2018 / Notices

                                               measurements met the minimum                               For the ITCP weld of DSC 16, the                   representative of those that may be
                                               requirements of 0.5 inches for the OTCP                 applicant provided a calculation,                     found on DSCs 11–15.
                                               weld for DSCs 11–15.                                    AREVA Calculation 11042–0204,                            For the OTCP, the applicant stated
                                                  Based on the review of the                           Revision 3, ‘‘Allowable Flaw Size                     SIA concluded that the defects located
                                               information provided by the applicant,                  Evaluation in the Inner Top Cover Plate               within the weld deposit of DSC 16 are
                                               the NRC staff determined that the                       Closure Weld for DSC #16’’ (Exemption                 believed to be inter-bead lack of fusion
                                               required tests were performed on the                    Request Enclosure 4) that documents                   formed at the interface between adjacent
                                               ITCP and OTCP welds including in-                       the critical flaw size based on the                   weld bead surfaces. The applicant stated
                                               process visual inspections of welds                     maximum radial stresses in the welds                  that when the defects are present in the
                                               performed by the welders, VT of fit-ups                 due to design loads. The applicant’s                  DSC OTCP closure weld, they would be
                                               and welds and the vacuum hold, as well                  analysis calculated the critical flaw size            found at the interfaces between weld
                                               as helium pressure and helium leak                      for a weld size of 0.25 inch per the                  beads. The applicant included a
                                               testing. The NRC staff determined that                  PAUT results for DSC 16, which showed                 schematic showing the DSC OTCP weld
                                               the applicant completed an adequate                     that the distance between the weld root               bead placement and the position of the
                                               extent of condition assessment which                    and crown at the canister wall for the                lack-of-fusion flaws, which were
                                               showed that the welding of the ITCP                     DSC 16 ITCP lid weld ranged from 0.25                 characterized as parallel and offset. The
                                               and OTCP were conducted in                              inch to 0.4 inch. The applicant                       applicant stated that the possible
                                               accordance with welding administrative                  determined that the critical flaw depth               locations where lack of fusion between
                                               requirements, the required testing of                   was 0.15 inch, which would exceed the                 the sides of adjacent weld beads could
                                               welds were in compliance with                           typical weld layer thickness. The                     form in the DSC OTCP closure weld
                                               technical specifications, and weld depth                applicant noted that the measured weld                would result in defects that are not
                                               measurements for the OTCP met design                    size for the ITCP weld on DSC 16 was                  aligned and which would not extend
                                               requirements for the 61BTH DSC.                         significantly larger than the design                  beyond the thickness of one weld pass
                                               Adequate Strain Margins to                              thickness of 3/16 inch (i.e., 0.188’’). The           layer.
                                               Accommodate Flaws (Exemption                            applicant stated that all analyses for                   For the ITCP, the applicant stated SIA
                                               Request Enclosures 2 through 5): The                    DSCs 11–15 were conducted using the                   concluded that the locations of the flaws
                                                                                                       design thickness of the weld. The                     in DSC 16 indicate that they were
                                               applicant stated that strain margins for
                                                                                                       applicant provided an analysis of the                 related to sidewall lack of fusion. SIA
                                               DSCs 11–15 were demonstrated by
                                                                                                       allowable flaw size for the DSC ITCP                  also noted that the weld joint geometry,
                                               structural analysis using theoretically-
                                                                                                       and OTCP using the weld design                        welding system, and welding setup for
                                               bounding full-circumferential flaws and
                                                                                                       thickness which used the flaw sizes                   the ITCP of DSCs 11–15 had potential
                                               a structural analysis assuming flaw
                                                                                                                                                             for forming defects on the sidewall like
                                               distributions conservatively derived                    from the PAUT examination of DSC–16
                                                                                                                                                             those identified in DSC 16. The
                                               from the PAUT examination of DSC 16.                    (Exemption Request Enclosure 5,
                                                                                                                                                             applicant stated that, from the review,
                                               The applicant supported the analysis                    AREVA Calculation 11042–0205,
                                                                                                                                                             SIA concluded the other five canister
                                               using:                                                  Revision 3, ‘‘61BTH ITCP and OTCP
                                                                                                                                                             ITCP closure welds were welded in a
                                                  • A review of weld head video for all                Closure Weld Flaw Evaluation’’).
                                                                                                                                                             similar manner, using similar welding
                                               available DSCs, general area video for                     The applicant stated that, as part of              procedures, equipment, welding
                                               all available DSCs, and welding records;                the original extent of condition review,              process, filler material, and welding
                                                  • the allowable flaw size evaluation                 weld head videos were reviewed by SIA                 operators and thus, it is reasonable to
                                               in the ITCP closure weld for DSC 16;                    in 2014. For DSCs 13 and 16, the review               assume the other canister ITCP welds
                                               and                                                     included video recordings of the ITCP                 will have similar intermittent defects. In
                                                  • the ITCP and OTCP closure weld                     root and cover weld layers and the                    addition, the applicant stated that the
                                               flaw evaluation for a 61BTH DSC based                   OTCP tack, root, intermediate and cover               vertical weld wall of the weld groove is
                                               on the DSC 16 PAUT results.                             weld layers. For DSCs 12, 14 and 15, the              inherent to a single bevel design, and
                                                  Based on the review of the videos,                   review included video recordings of the               because there is limited room to tilt the
                                               welding records and the PAUT                            OTCP tack, root, intermediate and cover               tungsten electrode towards the side wall
                                               examination of DSC 16, the applicant                    weld layers. The applicant stated that                (DSC shell), any lack-of-fusion defects
                                               determined that the indications found                   no weld head video was available for                  that might form would likely be located
                                               on DSC 16 are representative of those                   DSC 11. The DSC 16 outer closure weld                 on the vertical sidewall. The applicant
                                               that may be found on DSCs 11–15.                        was concluded to be the most                          concluded that the assumptions made
                                               Consequently, the applicant determined                  vulnerable to potential defects because               for the ITCP closure weld bounding
                                               that the same bounding analyses                         a greater frequency of irregular surface              analysis in DSC 16 were considered
                                               performed for DSC 16 should provide                     conditions was generated during                       reasonable for all ITCP canister closure
                                               for similar conservative results for the                welding.                                              welds.
                                               closure welds for DSCs 11–15. The                          The applicant stated that SIA                         The NRC staff reviewed the
                                               applicant stated that for the OTCP, the                 performed further reviews of available                applicant’s summary of the weld head
                                               original design basis calculations                      weld head videos along with general                   video and general area videos. The NRC
                                               determined critical flaw sizes. The                     area videos, welding records, and PAUT                staff also reviewed the applicant’s
                                               applicant stated that these design basis                results for DSC 16 to identify any                    supporting analyses including:
                                               analyses determined for a 360°                          correlations between the welding                         • AREVA Calculation 11042–0204,
                                               circumferential flaw, an allowable flaw                 processes used during the 2013 loading                Revision 3, ‘‘Allowable Flaw Size
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                                               depth of 0.19 inch and 0.29 inch could                  campaign and the flaws identified by                  Evaluation in the Inner Top Cover Plate
                                               exist for surface connected and sub-                    the PAUT. The applicant stated that, by               Closure Weld for DSC #16’’ (Exemption
                                               surface flaws respectively. Finally, the                correlating indications to the particular             Request Enclosure 4);
                                               applicant stated that the flaw sizes                    welding methods used on all six                          • AREVA Calculation 11042–0205,
                                               determined by these calculations bound                  canisters (including DSCs 11–15), a                   Revision 3, ‘‘61BTH ITCP and OTCP
                                               any of the indications found on DSC 16                  reasonable case was made that the types               Closure Weld Flaw Evaluation’’
                                               by PAUT of the OTCP weld.                               of indications found on DSC 16 are                    (Exemption Request Enclosure 5);


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                                                                         Federal Register / Vol. 83, No. 181 / Tuesday, September 18, 2018 / Notices                                           47197

                                                  • Structural Integrity Associates, Inc.              unlikely that a connected lack-of-fusion              ITCP weld and the ITCP or the interface
                                               Report 700388.401, Revision 1,                          defect greater than the thickness of one              of the ITCP weld and the DSC shell.
                                               ‘‘Evaluation of the Welds on DSC 11–                    pass would be present. The NRC staff                  Given the geometry of the weld joint,
                                               15’’ (Exemption Request Enclosure 3);                   determined that any lack-of-fusion                    the number of welding passes required
                                                  • Structural Integrity Associates Inc.               defects in the OTCP would not be                      to fill the weld joint, the position of
                                               Report 1301415.403, Revision 2,                         aligned because of the weld joint                     each welding pass, and the requirement
                                               ‘‘Assessment of Monticello Spent Fuel                   geometry and the positioning of the                   for in-process visual inspection of the
                                               Canister Closure Plate Welds Based on                   weld passes required to fill the OTCP                 weld after each pass, the NRC staff
                                               Welding Video Records’’ dated May 22,                   weld joint.                                           determined that lack of fusion between
                                               2014 (RAI Response 1 Enclosure 8);                         With respect to the ITCP welds, the                the ITCP weld and the DSC shell is
                                                  • Structural Integrity Associates Inc.               NRC staff reviewed the applicant’s                    likely to be the most significant type of
                                               Report 1301415.402, Revision 0,                         analysis for the ITCP welds and the                   weld defect in this joint. The NRC staff
                                               ‘‘Review of TRIVIS Inc. Welding                         description of the ITCP welding based                 determined that the positioning of the
                                               Procedures used for Field Welds on The                  on weld head video described in                       welding electrode necessary to weld the
                                               Transnuclear NUHOMS® 61BTH Type 1                       Exemption Request Enclosure 3,                        root pass would minimize the chances
                                               & 2 Transportable Canister for BWR                      Structural Integrity Associates, Inc.                 of a lack-of-fusion defect located at the
                                               Fuel’’ (RAI Response 1 Enclosure 9);                    Report 700388.401, Revision 1,                        interface of the ITCP weld and the ITCP.
                                               and                                                     ‘‘Evaluation of the Welds on DSC 11–                  The NRC staff determined that the
                                                  • RAI Response 2.                                    15.’’ The NRC staff also reviewed the                 positioning of the welding electrode
                                                  The NRC staff determined that,                       following appendices to Exemption                     necessary to weld the second fill pass
                                               because the same welding process,                       Request Enclosure 3: Appendix A,                      would minimize the chances of a lack-
                                               welding equipment, and welding                          ‘‘Inner Top Cover Plate Closure Weld                  of-fusion defect at the interface of the
                                               procedures were used by the personnel                   Bead Sequence (Based on VID                           ITCP weld and the DSC shell.
                                               that conducted the ITCP and OTCP                        Observations)’’; Appendix C,                             Based on the review of the
                                               welds in DSCs 11–16, it is reasonable to                ‘‘Tabulated Review of Available VIDS                  information provided by the applicant
                                               conclude, based on engineering                          for Monticello DSC–12 through DSC–                    including the review of weld head video
                                               judgement that the types of defects in                  16’’; and Appendix D ‘‘Monticello DSC                 for all available DSCs, general area
                                               DSC 16 are representative of those that                 Video Inspection.’’                                   video for all available DSCs, and
                                               may be in DSCs 11–15. The NRC staff                        The NRC staff notes that it is unclear             welding records; the allowable flaw size
                                               determined that, because the DSCs 11–                   whether some of the observations in                   evaluation in the ITCP closure weld for
                                               16 are the same design, were fabricated                 Exemption Request Enclosure 3,                        DSC 16; and the ITCP and OTCP closure
                                               to the same specifications, and were                    Appendix C were in conformance with                   weld flaw evaluation for a 61BTH DSC
                                               subjected to the same tests, the analysis               Procedure 12751–MNGP–OPS–01,                          based on the DSC 16 PAUT results, the
                                               conducted for DSC 16 is also applicable                 Revision 0, ‘‘Spent Fuel Cask Welding:                NRC staff concludes that the applicant
                                               to DSCs 11–15.                                          61BT/BTH NUHOMS® Canisters’’ (RAI                     has adequately considered the sizes and
                                                  The NRC staff reviewed the                           Response 1 Enclosure 6). In particular,               location of potential weld flaws to
                                               applicant’s analysis for the OTCP welds                 the NRC staff note that Exemption                     evaluate the stress margins in the ITCP
                                               and the description of the OTCP                         Request Enclosure 3, Appendix C                       and OTCP welds of DSCs 11–15. The
                                               welding based on weld head video                        indicated there were two instances of                 NRC staff structural review for the
                                               described in Exemption Request                          blow through of the root pass on the                  requested exemption follows the
                                               Enclosure 3, Structural Integrity                       OTCP weld of DSC–12. Procedure                        materials review.
                                               Associates, Inc. Report 700388.401,                     12751–MNGP–OPS–01, Revision 0                            Additional Strain Margins in Welds
                                               Revision 1, ‘‘Evaluation of the Welds on                states such an event would be treated as              (Exemption Request Enclosures 6
                                               DSC 11–15,’’ Appendix B, ‘‘Outer Top                    a major repair with additional NDE and                through 9): The applicant stated that
                                               Cover Plate Closure Weld Bead                           documentation. However, in RAI                        additional analysis was performed to
                                               Sequence (Based on VID Observations)’’                  Response 2, the applicant indicated that              maximize the size of flaws present in
                                               and Appendix C, ‘‘Tabulated Review of                   these events were weld craters and were               locations consistent with the results of
                                               Available VIDS for Monticello DSC–12                    not weld root blow through events.                    the DSC 16 PAUT to demonstrate
                                               thru DSC–16.’’ The NRC staff also                       While NRC staff was not able to resolve               substantial margin to account for
                                               reviewed the information included from                  whether these actions taken by the                    potential flaw uncertainties. In addition,
                                               the review of the general area video                    welder were in conformance with the                   the applicant stated that DSCs 11–15
                                               records included in Appendix D of                       applicable procedure, it was apparent                 site-specific heat load conditions were
                                               Exemption Request Enclosure 3,                          from Exemption Request Enclosure 3,                   applied to demonstrate additional weld
                                               ‘‘Monticello DSC Video Inspection.’’                    Appendix C that corrective actions were               margin exists and is available to account
                                               The NRC staff determined that due to                    taken to address the weld defects. In                 for any remaining flaw uncertainty. The
                                               the OTCP weld joint design and welding                  addition, the NRC staff determined that               applicant stated that the analysis used
                                               process used in the OTCP closure weld,                  either a blow through of the root pass                design basis loads with flaws present in
                                               the likely significant welding defects in               or a weld crater is a localized defect that           locations consistent with the DSC 16
                                               the OTCP weld would be lack of fusion                   would, in the worst case, compromise a                PAUT results and maximized in size
                                               between the weld beads or at the                        small length of the root pass. As such,               such that the weld flaws approach
                                               interface of the OTCP weld and the                      the NRC staff determined that the                     acceptable design limits.
                                               OTCP or the interface of the OTCP weld                  reported observation of a possible root                  The applicant stated that the two
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                                               and the DSC shell. Given the geometry                   blow through in two locations is bound                maximum modeled weld flaws for
                                               of the weld joint, the number of welding                by the assumed size of the OTCP welds                 OTCP to DSC shell weld are 0.43 inch
                                               passes required to fill the weld joint, the             defects in the flaw evaluation.                       and 0.42 inch in height, which
                                               position of each welding pass, and the                     The NRC staff determined that for the              represents about 85% through-wall of
                                               requirement for in-process visual                       ITCP weld joint design the likely                     the 0.5-inch minimum weld throat. The
                                               inspection of the weld after each pass,                 significant welding defects would be                  applicant stated that the maximum
                                               the NRC staff determined that it is                     lack of fusion at the interface of the                modeled full-circumferential weld flaws


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                                               47198                     Federal Register / Vol. 83, No. 181 / Tuesday, September 18, 2018 / Notices

                                               for ITCP to DSC shell weld are 0.11 inch                HSMs. As noted by the applicant, one                  to conservatively represent the closure
                                               in height at the ITCP weld to the ITCP                  of the many factors contributing to this              weld flaws for DSCs 11–15. As noted in
                                               interface and 0.14 inch in height at the                assertion is the structural integrity of the          the Materials Review, the NRC staff
                                               ITCP weld to DSC shell interface, which                 DSC top cover plates-to-shell closure                 reviewed the applicant’s evaluation and
                                               represent respectively 58% and 74%                      welds. The Structural Review is based                 determined that the flaws used in
                                               through-wall of the 0.19-inch minimum                   on the conclusion of the Materials                    analyzing the DSC 16 closure welds are
                                               weld throat. The applicant stated that                  Review where the NRC staff determined                 a reasonable representation for the
                                               each of the four assumed flaws                          among other findings that, because the                closure welds for all DSCs 11–16. This
                                               represent defects spreading over more                   DSCs 11–16 are of the same design,                    finding provides the basis for the NRC
                                               than one weld bead.                                     were fabricated to the same                           staff to review the two calculation
                                                  The NRC staff reviewed the                           specifications, and were subjected to the             packages: Calculations 11042–0207 and
                                               applicant’s analysis for the ITCP and                   same tests, the analyses conducted for                11042–0208, which used the maximized
                                               OTCP weld flaws along with the                          DSC 16 may also be applied to DSCs 11–                weld flaws that are essentially the same
                                               applicant’s summary of the welding                      15.                                                   in distribution but are much larger in
                                               video recordings and the PAUT                              For the DSC 11–15 closure weld                     size than those used for the DSC 16
                                               examination results for DSC 16. For the                 structural functions assessment, which                evaluation.
                                               ITCP weld, the NRC staff assessed the                   was done by analysis, the applicant                      Specifically, in Calculation 11042–
                                               geometry of the weld joint, the                         noted that the previous evaluations to                0207, the applicant asserts that there are
                                               positioning of the welding electrode in                 demonstrate adequate strain margins of                adequate strain margins in the welds to
                                               both the root and the final fill pass along             safety of the DSC 16 closure welds also               accommodate flaws for DSCs 11–15.
                                               with the requirement for in-process                     support the current exemption request.                The DSCs are subject to the design basis
                                               visual inspection of the weld after each                These evaluations were provided in the                temperature, pressure, and side-drop
                                               pass. For the OTCP weld, the NRC staff                  following reports:                                    loading conditions and are analyzed per
                                               assessed the geometry of the weld joint,                   • SIA Report 1301415.301, Revision                 the ASME Code Section III criteria,
                                               the number of welding passes required                   0, ‘‘Development of an Analysis Based                 using the limit load and elastic-plastic
                                               to fill the weld joint, the position of                 Stress Allowable Reduction Factor                     analyses. In Calculation 11042–0208,
                                               each welding pass, along with the                       (SARF)—Dry Shielded Canister (DSC)                    the applicant asserts additional strain
                                               requirement for in-process visual                       Top Closure Weldments’’ (Exemption                    margin in the DSCs 11–15 closure
                                               inspection of the weld after each pass.                 Request Enclosure 2);                                 welds. The maximum flaws, the
                                               The NRC staff determined that any lack-                    • AREVA Calculation 11042–0204,                    analysis methodology and the
                                               of-fusion defects in the ITCP and OTCP                  Revision 3, ‘‘Allowable Flaw Size                     evaluation criteria are the same as those
                                               would not be aligned and would not                      Evaluation in the Inner Top Cover Plate               of Calculation 11042–0207. However, in
                                               result in a defect greater than the                     Closure Weld for DSC #16’’ (Exemption                 lieu of the design basis loading, the
                                               thickness of one pass given the weld                    Request Enclosure 4); and                             analysis used the as-loaded DSC cavity
                                               joint geometry and the positioning of                      • AREVA Calculation 11042–0205,                    pressure, which is site-specific and
                                               the weld passes required to fill the ITCP               Revision 3, ‘‘61BTH ITCP and OTCP                     temperature dependent. The at-
                                               and OTCP weld joints. Thus, the NRC                     Closure Weld Flaw Evaluation’’                        temperature material yield strengths are
                                               staff determined that the flaws assessed                (Exemption Request Enclosure 5).                      used, which are higher than those
                                               in Exemption Request Enclosure 6 are                       The evaluations performed on the                   associated with the design basis loading.
                                               both unlikely to occur in any of the                    DSC 16 closure welds included: (1) A                     It is noted that the exemption request
                                               DSCs loaded in the 2013 campaign and                    structural analysis using an analysis-                also included Calculation 11042–0209
                                               the flaws assessed in Exemption                         based stress allowance reduction factor               (Exemption Request Enclosure 8) to
                                               Request Enclosure 6 conservatively                      and theoretically-bounding full-                      demonstrate additional weld strain
                                               bound any possible welding defects that                 circumferential flaws to demonstrate                  margin for DSCs 11–15 subject to the
                                               are likely to exist in the DSC 11–15                    that finite element analysis (FEA)                    site-specific side-drop loading
                                               OTCP welds.                                             simulation is suitable for analyzing the              condition. The NRC staff neither
                                                  Based on the review of the                           structural performance of the weld as a               approves, nor rejects, and is not
                                               information provided by the applicant                   continuum with multiple embedded                      expressing any view related to the
                                               including the analysis of flaws analyzed                flaws; (2) a calculation that documents               material in the calculation, as it did not
                                               from the PAUT examination of the ITCP                   the allowable critical flaw size in the               enter into the NRC evaluation.
                                               and OTCP welds of DSC 16 and the                        ITCP closure weld based on the                           The NRC staff reviewed the above two
                                               assumed maximized flaws that exceed                     maximum design basis radial stresses in               calculation reports on the structural
                                               the weld bead deposit thickness, the                    the welds; and (3) a structural analysis              performance of the DSC 11–15 closure
                                               NRC staff concludes that the applicant’s                demonstrating large weld strain margins               welds. In Calculation 11042–0207, the
                                               analysis of stress margins in the ITCP                  of safety with conservative assumptions               applicant followed the same analysis
                                               and OTCP welds of DSCs 11–15                            of flaw distribution and size derived                 method used in Calculation 11042–0205
                                               conservatively assumed weld flaws that                  from the DSC 16 PAUT examination                      for DSC 16 to demonstrate adequate
                                               are much larger than would be                           results.                                              strain margin in DSCs 11–15 closure
                                               reasonably expected. This is due to the                    However, to demonstrate adequate                   welds. The applicant noted that the
                                               combination of the materials of                         strain margin and to accommodate flaws                finite element model details and
                                               construction, weld joint designs, and                   in the DSCs 11–15 closure welds, the                  structural performance acceptance
                                               the welding process used for the ITCP                   applicant provides a FEA simulation                   criteria are the same except that the
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                                               and OTCP welds.                                         evaluation in SIA Report, 700388.401,                 maximized flaw configuration is
                                                  Structural Review for the Requested                  Revision 1, ‘‘Evaluation of the Welds on              postulated to result in much larger flaws
                                               Exemption: The exemption request                        DSCs 11–15,’’ (Exemption Request                      than those associated with DSC 16 to
                                               states that there is a reasonable                       Enclosure 3) to support that the flaw                 provide additional insights into the
                                               assurance of safety to grant the                        distribution and size based on the PAUT               weld structural performance.
                                               requested exemption to continue the                     examination results for the DSC 16                       To arrive at the maximized
                                               storage of DSCs in their respective                     closure weld performance can be used                  configuration, the flaws modeled in


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                                                                         Federal Register / Vol. 83, No. 181 / Tuesday, September 18, 2018 / Notices                                          47199

                                               Calculation 11042–0205 for DSC 16                       limit load analysis. The applicant                    positive, even in loading conditions that
                                               were first modified slightly, including                 determined the side-drop collapse load                are 50 percent beyond those required for
                                               replacing conservatively the 0.11 inch-                 to be approximately 179.5 g, which                    evaluating localized strains by the
                                               long flaw inside the ITCP with an                       includes an off-normal DSC design basis               elastic-plastic analysis, the NRC staff
                                               equivalent-height flaw at the interface                 internal pressure of 20 psi as a boundary             concludes that there are adequate strain
                                               between the ITCP and the 3/16-inch                      condition. This determination is                      margins on the welds to accommodate
                                               ITCP-to-shell weld. However, the size                   acceptable because the collapse load is               flaws for DSCs 11–15.
                                               and location of all other welds were                    greater than the required side-drop load                 The applicant noted that there are
                                               unchanged. Next, an elastic-plastic                     of 104 g to satisfy the ASME Code limit-              additional strain margins in the closure
                                               analysis of flaw length introduced                      load analysis acceptance criteria.                    welds of DSCs 11–15 owing to the site-
                                               increasingly larger flaw sizes in each                     To address the potential material                  specific as-loaded temperature and DSC
                                               analysis iteration to simulate higher                   rupture associated with high plastic                  internal pressure conditions at MNGP,
                                               localized plastic strain. As noted by the               strain concentrations at the weld flaws,              which are less severe than those
                                               applicant, the iteration analysis was                   the applicant performed elastic-plastic               associated with the design basis
                                               considered complete for the maximized                   analyses in Calculation 11042–0207 to                 conditions. In Calculation 11042–0208
                                               flaws determination for which the peak                  quantify strain margins of safety for the             (Exemption Request Enclosure 7), the
                                               equivalent plastic strain for the most                  DSCs 11–15 with maximized flaws. This                 applicant performed evaluations using
                                               critically stressed flaws would be                      concern was addressed by considering a                the temperature and pressure conditions
                                               calculated to be somewhat below the                     Ramberg-Osgood idealization of the                    specific to DSCs 11–15. The evaluation
                                               ASME code weld material elongation                      stress-strain curve for SA–240 Type 301               follows the same Calculation 11042–
                                               limit of 28 percent. The applicant                      stainless steel, which recognizes strain              0207 analysis method and acceptance
                                               performed the elastic-plastic iteration                 hardening effects for the FEA modeling.               criteria, including the same maximized
                                               analysis using a 150-percent design                     The elastic-plastic analyses resulted in              flaws. The applicant indicated that the
                                               basis side-drop of 112.5 g (75 × 1.5 =                  the peak equivalent plastic strains of 7.4            evaluations were intended to address
                                               112.5) to arrive at the maximized flaws.                percent and 11.1 percent for the Service              any remaining uncertainties related to
                                               Specifically, the maximized, 360° full-                 Level D design basis pressure of 65 psi               potential flaws that may be present in
                                               circumferential flaws are of 0.43 inch                  and side-drop of 75 g, respectively. For              DSCs 11–15 by demonstrating existence
                                               and 0.42 inch in height for the two flaws               the strain margin evaluation, the                     of additional strain margins in the
                                               associated with the OTCP, which                         applicant continued to use the same                   closure welds.
                                               represent about 85% through-wall of the                 DSC 16 weld strain acceptance criterion                  Using the site-specific 370 °F at-
                                               0.5-inch minimum throat for OTCP-to-                    of not exceeding the 28 percent                       temperature material yield strength of
                                                                                                       elongation limit, which is a reduction                21.2 ksi for the SA–240 Type 304
                                               DSC shell weld. The maximized full-
                                                                                                       from the ASME B&PV Code specified                     stainless steel, the applicant determined
                                               circumferential flaws for ITCP-to-DSC
                                                                                                       weld elongation limit of 35 percent by                the Service Level D limit load collapse
                                               shell weld are 0.11 inch and 0.14 inch
                                               each in height, which represent                         a factor of 0.8 (0.35 × 0.8 = 0.28).                  pressure is 144.1 psi. This pressure is
                                                                                                       Considering the 28 percent elongation                 significantly higher than the DSC at-
                                               respectively 58% and 74% through-wall
                                                                                                       limit, the strain margins of safety                   temperature internal pressure of 45.9 psi
                                               of the 0.19-inch minimum weld throat.
                                                                                                       corresponding to the calculated peak                  and the ASME Code limit-load collapse
                                               The NRC staff reviewed the iteration
                                                                                                       equivalent plastic strains are 2.78                   pressure acceptance criteria of 90.2 psi.
                                               analysis for arriving at the maximized
                                                                                                       {(0.28/0.074)¥1 = 2.78} and 1.52                      Correspondingly, using the site-specific
                                               flaws for the DSCs 11–15 closure welds.
                                                                                                       {(0.28/0.111)¥1 = 1.52}, respectively.                237 °F at-temperature material yield
                                               Because the maximized flaws are                                                                               strength of 24.0 ksi, together with the
                                                                                                       Because the margins of safety are all
                                               essentially the same in locations as                                                                          off-normal at-temperature internal
                                                                                                       positive (i.e., greater than zero), the NRC
                                               those used for DSC 16 and the resulting                                                                       pressure of 10.9 psi as a boundary
                                                                                                       staff concludes that there are adequate
                                               flaw sizes are much larger than the                                                                           condition, the applicant determined the
                                                                                                       strain margins in the welds to
                                               corresponding ones used for DSC 16, the                                                                       collapse side-drop g-load to be 204 g.
                                                                                                       accommodate flaws for DSCs 11–15.
                                               NRC staff concludes that the postulated                    Additionally, similar to the analysis              This site-specific collapse side-drop is
                                               maximized flaws are conservative and                    used to supplement qualification of the               also much greater than the ASME Code
                                               appropriate for evaluating the strain                   DSC 16 closure welds, the applicant                   limit-load collapse side-drop g-load
                                               performance of the DSCs 11–15 closure                   considered a 150 percent of the design                acceptance criteria of 104 g associated
                                               welds.                                                  basis loading to evaluate the DSCs 11–                with the design basis 500 °F at-
                                                  Using the maximized flaws, the                       15 welds. The analysis used a DSC                     temperature material yield strength of
                                               applicant performed limit load analyses                 internal pressure of 100 psi (65 × 1.5 =              19.4 ksi.
                                               in Calculation 11042–0207 for two DSC                   97.5 <100 psi) and a side-drop of 112.5                  To determine the strain margins of
                                               design basis internal pressures of 32 psi               g (75 × 1.5 = 112.5 g), which are beyond              safety for the site-specific temperature
                                               and 65 psi for the ASME Code Service                    the ASME B&PV Code, Section III,                      and pressure, the applicant performed
                                               Level A/B and Service Level D                           Paragraph NB–3228.3 Plastic Analysis                  elastic-plastic analyses for DSCs 11–15
                                               evaluations, respectively. The analyses                 provisions. The calculated peak                       with the maximized flaws in the OTCP-
                                               resulted in the calculated collapse                     equivalent plastic strains are 13.6                   and ITOP-to-shell welds. Using the
                                               pressures of 86.3 psi for Service Level                 percent and 23.0 percent for the                      analysis approach in Calculation 11042–
                                               A/B and 122.2 psi for Service Level D.                  respective pressure and side-drop                     0207, the applicant calculated the peak
                                               The collapse pressures are acceptable                   loading cases. For the weld strain                    equivalent plastic strains of 4.4 percent
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                                               because they are greater than the                       margin evaluation, the applicant                      and 9.8 percent for the Service Level D
                                               respective ASME Code limit-load                         continued to use the same 28 percent                  internal pressure of 45.9 psi and the
                                               analysis acceptance criteria of 60 psi                  weld elongation limit which resulted in               design basis side-drop of 75 g,
                                               and 90.2 psi. Similarly, for the design                 the weld strain margins of safety of 1.06             respectively. For the same weld
                                               basis DSC side-drop of 75 g, the                        {(0.28/0.0136)¥1 = 1.06} and 0.22                     elongation limit of 28 percent, the
                                               applicant used the 3D half-symmetric                    {(0.28/0.23)¥1 = 0.22}, respectively.                 corresponding strain margins of safety
                                               model to perform a Service Level D                      Because all margins of safety are                     are calculated to be 5.36 {(0.28/


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                                               47200                     Federal Register / Vol. 83, No. 181 / Tuesday, September 18, 2018 / Notices

                                               0.044)¥1 = 5.36} and 1.86 {(0.28/                       support the thermal design function by                remains in compliance with 10 CFR part
                                               0.098)¥1 = 1.86}. Similar to the                        virtue of their confinement function (as              72.
                                               analysis used in Calculation 11042–                     demonstrated by the helium leak test                     Shielding and Criticality Safety
                                               0207 for a supplement qualification of                  conducted on DSCs 11–15) which                        Review for the Requested Exemption:
                                               the DSC 16 closure welds with a more                    assures the helium atmosphere in the                  The NRC staff reviewed the criticality
                                               conservative loading assumption, the                    DSCs 11–15 cavity is maintained in                    safety and radiation protection
                                               applicant also considered 150 percent of                order to support heat transfer. The                   effectiveness of DSCs 11–15 presented
                                               the site-specific loading to evaluate the               applicant also stated that the                        in the applicant’s exemption request.
                                               weld flaws using a DSC internal                         satisfactory completion of two required               The NRC staff finds that the criticality
                                               pressure of 69 psi (45.9 × 1.5 = 69 psi)                vacuum pump-downs conducted on the                    safety and radiation protection of DSCs
                                               and side-drop load of 112.5 g. The                      DSCs demonstrated weld integrity of the               11–15 are not affected by the
                                               resulting peak equivalent plastic strains               ITCP confinement boundary. These                      nonconforming PT examinations for the
                                               are 7.1 percent and 19.0 percent, which                 pump-downs establish a differential                   following reasons: (1) The interior of
                                               correspond to the strain margins of                     pressure across the ITCP and siphon/                  DSCs 11–15 will continue to prevent
                                               safety of 2.94 {(0.28/0.071)¥1 = 2.94}                  vent block welds of approximately one                 water in-leakage which means that the
                                               and 0.47 {(0.28/0.19)¥1 = 0.47},                        atmosphere, which exceeds the                         system will remain subcritical under all
                                               respectively. For the MNGP site-specific                magnitude of the 10 psig design                       conditions; and (2) the nonconforming
                                               evaluation, because the margins of                      pressure used in stress analyses for                  PT examinations do not affect the
                                               safety are all positive, the NRC staff                  normal conditions. Although the                       radiation source term of the spent fuel
                                               concludes that the DSCs 11–15 weld                      vacuum pump-down imparts a pressure                   contents, or the configuration and
                                               strains have additional margins beyond                  differential in a reverse direction from              effectiveness of the shielding
                                               the design basis conditions.                            the confinement function, according to                components of the Standardized
                                                  On the basis of the review above, the                the applicant, the pump-down                          NUHOMS® system containing the
                                               NRC staff concludes that the limit load                 demonstrates the basic function of the                61BTH DSC, meaning that the radiation
                                               and elastic-plastic analysis results                    confinement boundary and the lack of a                protection performance of the system is
                                               showed that the welds would undergo                     through-weld flaw in the ITCP and                     not altered.
                                               localized plastic deformation. The                      siphon/vent block welds sufficient to                    The NRC staff finds that the criticality
                                               applicant’s evaluation indicated that no                cause a loss of cavity helium when in                 safety and shielding function of DSCs
                                               weld material rupture or breach of the                  service.                                              11–15, loaded under CoC No. 1004,
                                               DSCs 11–15 confinement boundary at                        The NRC staff reviewed the                          Amendment No. 10, addressed in the
                                               the closure welds is expected because of                applicant’s exemption request and also                exemption request remains in
                                               the adequate margins of safety against                  evaluated its effect on DSCs 11–15                    compliance with 10 CFR part 72.
                                               the weld elongation limits. For this                    thermal performance. The NRC staff                       Confinement Review for the
                                               reason, the NRC staff has reasonable                    concludes that the cask thermal                       Requested Exemption: The objective of
                                               assurance to conclude that the ITCP and                 performance is not affected by the                    the confinement evaluation was to
                                               OTCP welds of DSCs 11–15 have                           exemption request because the                         confirm that DSCs 11 through 15 loaded
                                               adequate structural margins of safety for               applicant has shown that a satisfactory               at the MNGP met the confinement-
                                               the ASME Code Service Level D design                    helium leak test was conducted on DSCs                related requirements described in 10
                                               criteria, which bound the normal, off-                  11–15, which is integral to ensuring                  CFR part 72. NRC staff relied on the
                                               normal, and accident (including natural                 integrity of the primary confinement                  information provided by the applicant
                                               phenomenon) conditions for the subject                  boundary. Integrity of the primary                    in their Exemption Request dated
                                               weld structural integrity evaluation. The               confinement boundary assures the spent                October 18, 2017.
                                               NRC staff also finds that the                           fuel is stored in a safe inert environment               As described in the applicant’s
                                               retrievability of DSCs 11–15 is ensured                 with unaffected heat transfer                         ‘‘Exemption Request for Nonconforming
                                               based on the demonstration of adequate                  characteristics that assure peak cladding             Dry Shielded Canister Dye Penetrant
                                               weld strain margins of safety discussed                 temperatures remain below allowable                   Examinations’’ (Exemption Request
                                               above.                                                  limits. The NRC staff also concludes                  Enclosure 1), certain elements of the
                                                  Thermal Review for the Requested                     that the applicant demonstrated the lack              DSCs 11–15 closure weld PT
                                               Exemption: The applicant stated that                    of a through-weld flaw in the ITCP and                examinations did not comply with
                                               even though nonconforming                               siphon/vent block weld sufficient to                  examination procedures associated with
                                               examinations exist for the primary                      cause a loss of cavity helium. This                   TS 1.2.5. To support the exemption
                                               confinement welds, satisfactory                         satisfies 10 CFR 72.236(f) which                      request, the applicant noted that a
                                               completion of the required helium leak                  requires that the cask be designed to                 helium leakage rate test of the closure’s
                                               test conducted on DSCs 11–15 has                        have adequate heat removal capacity                   confinement boundary, including ITCP
                                               demonstrated the integrity of the                       without active cooling systems and 10                 weld, siphon cover plate weld, and vent
                                               primary confinement boundary (ITCP                      CFR 72.122(h) which states that the fuel              port cover plate weld, were conducted
                                               and siphon/vent cover plate) welds.                     cladding during storage must be                       per TS 1.2.4a and demonstrated that the
                                               These tests specifically demonstrated                   protected against degradation and gross               primary confinement barrier field welds
                                               that the primary confinement boundary                   rupture. Therefore, based on the NRC                  met the TS acceptance criterion of
                                               field welds are ‘‘leak tight’’ as defined               staff’s review of the applicant’s                     leaktight as defined by ANSI N14.5–
                                               in ANSI N14.5–1997. The applicant                       evaluation and technical justification,               1997. The applicant noted that the
                                               stated that, in this respect, the helium                the NRC staff finds the exemption                     confinement integrity is not affected by
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                                               leak test demonstrated the basic                        request acceptable by virtue of the                   the non-compliant PT examination
                                               integrity of the primary confinement                    demonstrable structural integrity of the              procedures. The NRC staff concludes
                                               boundary and the lack of a through-                     ITCP and siphon/vent plate welds.                     that not performing the PT examination
                                               weld flaw in the field closure welds that                 The NRC staff finds that the thermal                procedures relevant to this exemption
                                               would lead to a loss of cavity helium in                function of DSCs 11–15, loaded under                  request would not change the results of
                                               DSCs 11–15. The applicant stated that                   CoC No. 1004, Amendment No. 10,                       the helium leakage test, which is
                                               the field closure welds indirectly                      addressed in the exemption request                    integral to ensuring closure confinement


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                                                                         Federal Register / Vol. 83, No. 181 / Tuesday, September 18, 2018 / Notices                                            47201

                                               integrity, and therefore, the closure                      The applicant applied the                          NUHOMS® DSC contains 61 BWR fuel
                                               confinement integrity is unaffected. The                methodology and results in NUREG–                     assemblies with 15.5-year-old fuel of 41
                                               structural and material acceptability of                1864 to perform the risk assessment.                  GWD/MTU (not high burnup) fuel. The
                                               DSCs 11 through 15 welds is discussed                   The risk assessment compared the                      plume heat content for a cask release is
                                               in the Structural Review and the                        NUHOMS® and HI–STORM–100 dry                          estimated to be that of the spent fuel.
                                               Materials Review described previously.                  spent fuel storage systems and                        NUREG–1864 estimates the maximum
                                                  It is noted that a dose-related analysis             determined the designs are similar with               decay heat load to be 264 watts per
                                               was included as Enclosure 10 of the                     a few basic differences. Both storage                 assembly. The estimated maximum
                                               Exemption Request. NRC staff neither                    systems include canisters for confining               decay heat load for MNGP DSC is
                                               approves, nor rejects, and is not                       dry spent fuel. The canisters have                    approximately 220 watts per assembly.
                                               expressing any view related to the                      similar design and dimensions and are                 The risk assessment analysis assumes
                                               material in that enclosure, as it did not               made of stainless steel of similar                    that the source term from NUREG–1864
                                               enter into the evaluation.                              thickness and are required to meet the                adequately represents or bounds those
                                                  Risk Assessment for the Requested                    same ASME class (ASME B&PV, Section                   of the MNGP configuration. The NRC
                                                                                                       III, and Subsection NB). The HI–STORM                 staff agrees that this is reasonable based
                                               Exemption: In support of the applicant’s
                                                                                                       100 system consists of a multipurpose                 on the applicant’s assessment which
                                               request, the applicant submitted a risk
                                                                                                       canister (MPC) that confines spent fuel               shows NUREG–1864 radionuclide
                                               assessment, Jensen Hughes Report
                                                                                                       assemblies, a transfer overpack that                  inventory is 7.0 times higher than that
                                               016045–RPT–01, ‘‘Risk Assessment of
                                                                                                       provides shielding during canister                    of MNGP DSC.
                                               MNGP DSCs 11–15 Welds Using                                                                                      The NUREG–1864 evaluation of
                                                                                                       preparation, and a vertical, cylindrical
                                               NUREG–1864 Methodology’’                                                                                      misload concluded MPC integrity would
                                                                                                       storage overpack that provides shielding
                                               (Exemption Request Enclosure 11). The                                                                         not be affected unless a gross series of
                                                                                                       during long-term storage.
                                               risk assessment compares the calculated                    Both MNGP and Hatch (the plant                     errors occurred. The errors would have
                                               risk of leaving the five DSCs in storage                selected for the Pilot PRA) are BWR,                  to result in nearly every fuel assembly
                                               ‘‘as is’’ at the MNGP ISFSI versus                      Mark 1 plants; therefore, the storage                 loaded into the MPC being incorrect and
                                               transferring the DSCs back into the                     systems are exposed to similar handling               insufficiently cooled. NUREG–1864
                                               reactor building to perform PAUT of the                 hazards. The potential drop heights for               concluded this gross misload scenario
                                               welds and then returning them to their                  loaded TCs moving across the refueling                was not credible. Therefore, the risk
                                               storage locations. The risk for each                    floor, or lowering from the height of                 assessment did not explore risk from
                                               potential accident, regardless of                       refueling floor to the ground floor of the            misloading of spent fuel.
                                               likelihood, can be generally summarized                 equipment hatch are very similar. The                    The applicant’s risk assessment
                                               by the following equation:                              potential impact surfaces are also                    assumes the annual risk for a DSC while
                                               Initiating Event Frequency (per Year) ×                 similar.                                              stored on the ISFSI would be the same
                                                    Probability of Canister Release ×                     The NUHOMS® system is comprised                    for both alternatives. The risk
                                                    Probability of Containment Release                 of a DSC, a TC, and an HSM. A transfer                assessment identified three types of
                                                    × Consequences (Cancer Fatality) =                 trailer is used to move the loaded TC.                mechanical failure that could cause
                                                    Risk                                               Two key differences exist between the                 significant radiological releases to the
                                                                                                       NUHOMS® and the HI–STORM dry                          environment: drop accidents, meteorite
                                               The process to transfer a DSC to the                    spent fuel storage operations. First, the             strikes, and overflight aircraft accidents.
                                               reactor building refueling floor for                    NUHOMS® TC is placed horizontally on                  The primary difference in risk between
                                               PAUT incurs added potential for                         the transfer trailer and is not subject to            the two alternatives, continued storage
                                               accidental drops due to the lifting and                 accidental drops when moving between                  at the ISFSI versus moving a DSC back
                                               subsequent lowering operations. For 20-                 the ISFSI and fuel building. Second,                  to the spent fuel pool area for PAUT, are
                                               year storage, the risk is the sum of all                transferring NUHOMS® DSC between                      potential drop accidents during lifting
                                               potential accident risks for the duration.              the TC and the HSM is done                            and lowering of a DSC between the
                                               Each DSC handling operation is                          horizontally; thus, the NUHOMS® DSC                   ground floor and the height of the
                                               independent. For five canisters, the total              is not subject to any potential vertical              refueling floor.
                                               risk value is multiplied by five.                       drop. During storage on an ISFSI pad,                    The applicant’s risk assessment
                                                  NUREG–1864, ‘‘A Pilot Probabilistic                  the horizontal-storage design of the                  accounted for possible added risk from
                                               Risk Assessment of a Dry Cask Storage                   HSM eliminates the risk of tip over                   a potential flaw around the canister lid
                                               System at a Nuclear Power Plant’’                       caused by seismic activities or wind-                 by assuming the probability of lid
                                               (ADAMS Accession No. ML071340012)                       driven missiles. Aircraft impact on the               failure would be same as for the DSC
                                               provides guidance for assessing the risk                HSM is limited to only large aircrafts                shell in drop accidents. This
                                               to the public and for identifying the                   and the methodology considered the                    assumption doubles the estimated
                                               dominant contributors to risk for                       distance to local airfields and planes                probability for a release from drop
                                               performing probabilistic risk                           that operate in the area. The NUREG–                  accidents. Strain analysis in NUREG–
                                               assessments (PRAs) of a dry cask storage                1864 frequency estimate for meteorite                 1864 reports the most highly stressed
                                               system located at a nuclear power plant                 strikes per unit area is used in this                 regions of the MPC for a drop accident
                                               site. NUREG–1864 documents a pilot                      assessment, and the analysis is adjusted              are in areas near the base of the
                                               PRA conducted for a dry cask storage                    for the larger horizontal surface area of             cylindrical shell and in the weld joining
                                               system (Holtec International HI–STORM                   the HSM.                                              the shell to the baseplate. Since the top
                                               100) at a Boiling Water Reactor (BWR)                      In the risk assessment, the potential              side of a canister is not expected to
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                                               Mark 1 plant. The risk assessment                       radiological consequences are based on                experience significant strain, the NRC
                                               estimated the annual off-site risk for one              a comparison of the spent fuel in the                 staff agrees that the assumption is
                                               cask in terms of individual probability                 MNGP DSC and the spent fuel modeled                   conservative and bounds the probability
                                               of a prompt fatality and a latent cancer                in NUREG–1864. In NUREG–1864, the                     of a release occurring following a drop
                                               fatality. It does not consider risk to                  HI–STORM 100 MPC contained 68 BWR                     accident.
                                               workers or future off-site transportation               fuel assemblies with 10-year-old high-                   The NRC staff reviewed the
                                               of DSCs.                                                burnup (50 GWD/MTU) fuel. The MNGP                    applicant’s risk assessment and agrees


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                                               47202                     Federal Register / Vol. 83, No. 181 / Tuesday, September 18, 2018 / Notices

                                               the mechanical failures identified and                  4.0 × 10¥2. This operation occurs inside              Assessment, Option 3 does not increase
                                               the radiological inventory from                         a closed building with probability of                 the risk by a discernible amount. All
                                               NUREG–1864 would be bounding for                        release value of 1.5 × 10¥4. The                      options would generate additional
                                               each of the MNGP DSCs. The risk                         consequence value for a release is 3.6 ×              radioactive contaminated material and
                                               assessment concludes that the risks are                 10¥4. The risk for a drop while lifting               waste from operations. For options 1
                                               significantly lower than the level                      a DSC up to the refueling floor can be                and 2, the lid would have to be
                                               considered ‘‘negligible’’ by the                        calculated as:                                        removed, which would generate
                                               Quantitative Health Guidelines (QHG)                    (5.6 × 10¥5)(4.0 × 10¥2)(1.5 × 10¥4)(3.6              cuttings from removing the weld
                                               established in ‘‘Risk-Informed                               × 10¥4) = 1.2 × 10¥13 cancer                     material that could require disposal as
                                               Decisionmaking for Nuclear Material                          fatality/year                                    contaminated material. For option 3,
                                               and Waste Applications,’’ Revision 1                    The risk for a drop while lowering a                  radioactive wastes would be generated
                                               (ADAMS Accession No. ML080720238).                      DSC (assuming no weld flaw,                           from radioactively contaminated
                                               The QHG considers public individual                     probability of release is 2.0 × 10¥2)                 consumables and anti-contamination
                                               risk of latent cancer fatality risk of less             through the equipment hatch back to                   clothing used during the examination.
                                               than 2 × 10¥6 per year as negligible. The               ground level can be calculated as:                    Also, radioactive waste would be
                                               pilot PRA (NUREG–1864) concluded                        (5.6 × 10¥5)(2.0 × 10¥2)(1.5 × 10¥4)(3.6              generated from the cleanup of any
                                               that there is no prompt fatality risk, and                   × 10¥4) = 6.0 × 10¥14 cancer                     coupling fluid (of the PAUT) that it
                                               the calculated risk is extremely small.                      fatality/year                                    combines with and then transports
                                               NUREG–1864 reports the increase in                                                                            resulting in contamination from the
                                                                                                       The additional risk from performing
                                               risk (individual probability of latent                                                                        surface of the DSC. This radioactive
                                                                                                       PAUT for five DSCs would be five times
                                               cancer fatality) from the first year as 1.8                                                                   waste would be transported and
                                                                                                       the sum of risk for lifting and lowering
                                               × 10¥12, and for subsequent years as 3.2                                                                      ultimately disposed of at a qualified
                                                                                                       one DSC.
                                               × 10¥14 per year per MPC. The total risk                                                                      low-level radioactive waste disposal
                                               for Monticello as calculated by Jensen                  5 × [(1.2 × 10¥13) + (6.0 × 10¥14)] = 9.3             facility, potentially exposing it to the
                                               Hughes took into account the                                 × 10¥13 cancer fatality/year                     environment.
                                               characteristics of the spent fuel and the               Probabilistic risk assessments are                       The proposed exemption to permit
                                               site, as well as the differences between                typically used to evaluate risks greater              continued storage of DSCs 11–15 in
                                               the MNGP and Hatch ISFSIs. For the                      than 1.0 × 10¥6. In light of the                      their respective HSMs for the service
                                               five DSCs over a period of 20-year                      calculated risk values, the NRC staff                 life of the canisters at the MNGP ISFSI
                                               storage, risk would be: Alternative 1,                  finds the off-site risk as too small to be            is consistent with NRC’s mission to
                                               continue storage as-is, Risk = 1.4 ×                    accurately discernable. Based on the                  protect public health and safety.
                                               10¥12; Alternative 2, move DSCs back                    discussion presented above, the NRC                   Approving the requested exemption
                                               up to the refueling floor for PAUT then                 staff concludes that risk to the public for           reduces the opportunity for a release of
                                               return to storage location, Risk = 2.3 ×                the two options provided by Jensen                    radioactive material compared to the
                                               10¥12; with a difference in risk between                Hughes, ‘‘continued storage as-is’’ and               alternatives to the proposed action,
                                               the two proposed alternatives of 9.3 ×                  ‘‘transfer, perform PAUT, and return to               because there will be no operations
                                               10¥13.                                                  storage,’’ are essentially equivalent.                involving the opening of the DSCs,
                                                  The assessment of difference in risk                 Otherwise in the Public Interest                      which confine the spent nuclear fuel,
                                               between the proposed alternatives was                                                                         and there will be no operations
                                               performed based on evaluation data                         In considering whether granting the                involving the opening of the HSMs
                                               from NUREG–1864. The MNGP off-site                      exemption is in the public interest, the              potentially exposing radioactive waste
                                               consequence is based on individual risk                 NRC staff considered the alternative of               to the environment. Therefore, the
                                               and not absolute population difference.                 not granting the exemption. If the                    exemption is in the public interest.
                                               Based on the considerations taken into                  exemption were not granted, in order to
                                                                                                       comply with the CoC, either (1) DSCs                  Environmental Consideration
                                               account for the difference between the
                                               NUREG–1864 MPC and the MNGP DSCs                        11–15 would have to be removed from                      The NRC staff also considered in the
                                               in this assessment, the NRC staff finds                 their respective HSMs, opened and                     review of this exemption request
                                               the risk assessment calculation to be                   unloaded, and the contents loaded in                  whether there would be any significant
                                               reasonable because the applicant used                   new DSCs, with each of those new DSCs                 environmental impacts associated with
                                               accepted methods and the site-specific                  welded and tested, or (2) removed from                the exemption. The NRC staff
                                               considerations were addressed in an                     the HSMs to allow access to the OTCP                  determined that this proposed action
                                               appropriately conservative manner.                      to be machined off, and the ITCP weld                 fits a category of actions that do not
                                                  The purpose of this assessment is to                 machined down to the root weld; and                   require an environmental assessment or
                                               compare the risk associated with leaving                each DSC, ITCP and OTCP inspected to                  environmental impact statement.
                                               these DSCs as-is at the ISFSI versus                    determine if there was any damage as a                Specifically, the exemption meets the
                                               transferring the five DSCs back to the                  result of the machining (which would                  categorical exclusion in 10 CFR
                                               refueling floor for PAUT, and then                      then necessitate the actions detailed in              51.22(c)(25).
                                               returning them to the ISFSI for storage.                option 1); or (3) conduct PAUT by                        Granting this exemption from 10 CFR
                                               The process of returning the five DSCs                  opening the HSMs to conduct in-situ                   72.212(a)(2), 72.212(b)(3),
                                               to the refueling floor for PAUT incurs                  testing (which is limited to less than                72.212(b)(5)(i), 72.214, and
                                               additional crane operation. The                         360° of the weld circumference) or                    72.212(b)(11) only relieves the applicant
                                               inadvertent drop frequency for heavy                    transferring to a TC for testing on the               from the inspection or surveillance
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                                               loads (NUREG–1774, ‘‘A Survey of                        ISFSI pad or in the reactor building                  requirements associated with
                                               Crane Operating Experience at U.S.                      (essentially Alternative 2 in the Risk                performing PT examinations with regard
                                               Nuclear Power Plants from 1968                          Assessment). Options 1 and 2 would                    to meeting TS 1.2.5 of Attachment A of
                                               through 2002’’, ADAMS Accession No.                     entail a higher risk of cask handling                 CoC No. 1004. A categorical exclusion
                                               ML032060160) is 5.6×10¥5/lift. The                      accidents, additional personnel                       for inspection or surveillance
                                               probability of release from a DSC drop                  exposure, and greater cost to the                     requirements is provided under 10 CFR
                                               accident, assuming defective weld, is                   applicant. As noted above in the Risk                 51.22(c)(25)(vi)(C) if the criteria in 10


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                                                                          Federal Register / Vol. 83, No. 181 / Tuesday, September 18, 2018 / Notices                                                       47203

                                               CFR 51.22(c)(25)(i)–(v) are also satisfied.              requested exemption neither changes                      radiological accidents such as a gross
                                               In its review of the exemption request,                  the effluents nor produces additional                    leak from the closure welds, because the
                                               the NRC staff determined, as discussed                   avenues of effluent release; (iii) granting              exemption neither reduces the ability of
                                               above, that, under 10 CFR 51.22(c)(25):                  the exemption would not result in a                      the closure welds to confine radioactive
                                               (i) Granting the exemption does not                      significant increase in either                           material nor creates new accident
                                               involve a significant hazards                            occupational radiation exposure or                       precursors at the MNGP ISFSI.
                                               considerations because granting the                      public radiation exposure, because the                   Accordingly, this exemption meets the
                                               exemption neither reduces a margin of                    requested exemption neither introduces                   criteria for a categorical exclusion in 10
                                               safety, creates a new or different kind of               new radiological hazards nor increases                   CFR 51.22(c)(25)(vi)(C).
                                               accident from any accident previously                    existing radiological hazards; (iv)
                                               evaluated, nor significantly increases                   granting the exemption would not result                  IV. Availability of Documents
                                               either the probability or consequences                   in a significant construction impact,
                                               of an accident previously evaluated; (ii)                because there are no construction                          The documents identified in the
                                               granting the exemption would not                         activities associated with the requested                 following table are available to
                                               produce a significant change in either                   exemption; and; (v) granting the                         interested persons through one or more
                                               the types or amounts of any effluents                    exemption would not increase either the                  of the following methods, as indicated.
                                               that may be released offsite because the                 potential or consequences from

                                                                                                                                                                                                          ADAMS
                                                                                                                   Document                                                                            accession No.

                                               Federal Register Notice Issuing Exemption from Nonconforming Dye Penetrant Examinations of Dry Shielded Canister                                        ML16159A227
                                                  (DSC) 16, June 8, 2016.
                                               Exemption Request for Nonconforming Dye Penetrant Examinations of Dry Shielded Canisters (DSCs) 11 through 15, October                                  ML17296A205
                                                  18, 2017.
                                               First Request for Additional Information for Review of Exemption Request for Five Nonconforming Dry Shielded Canisters 11                               ML18065A545
                                                  through 15 (CAC No. 001028, Docket No. 72–58, EPID L–2017–LLE–0029), March 6, 2018.
                                               Monticello Nuclear Generating Plant—Response to Request for Additional Information Regarding Exemption Request for Non-                                 ML18100A173
                                                  conforming Dye Penetrant Examinations of Dry Shielded Canisters (DSCs) 11 through 15, April 5, 2018.
                                               Supplement to Exemption Request for Nonconforming Dye Penetrant Examinations of Dry Shielded Canisters (DSCs) 11                                        ML18151A870
                                                  through 15 (CAC No. 001028, EPID L–2017–LLE–0029).
                                               NUREG–1774, ‘‘A Survey of Crane Operating Experience at U.S. Nuclear Power Plants from 1968 through 2002’’ .....................                        ML032060160
                                               Risk-Informed Decisionmaking for Nuclear Material and Waste Applications, Revision 1 ................................................................   ML080720238
                                               NUREG–1536, Revision 1 ‘‘Standard Review Plan for Spent Fuel Dry Storage Systems at a General License Facility’’ ...............                        ML101040620
                                               NUREG–1864, ‘‘A Pilot Probabilistic Risk Assessment of a Dry Cask Storage System at a Nuclear Power Plant’’ ........................                    ML071340012
                                               Attachment A, Technical Specifications, Transnuclear, Inc., Standardized NUHOMS® Horizontal Modular Storage System Cer-                                 ML17338A114
                                                  tificate of Compliance No. 1004, Renewed Amendment No. 10, Revision 1.



                                               V. Conclusion                                            NUCLEAR REGULATORY                                       Action to TS 3.8.4, Condition B, to
                                                                                                        COMMISSION                                               extend the completion time from 2
                                                  Based on the foregoing                                                                                         hours to 18 hours to repair each affected
                                               considerations, the NRC staff has                        [Docket No. 50–445; NRC–2018–0205]                       battery cell.
                                               determined that, pursuant to 10 CFR                                                                               DATES: Submit comments by October 2,
                                               72.7, the exemption is authorized by                     Vistra Operations Company LLC;
                                                                                                                                                                 2018. Requests for a hearing or petition
                                               law, will not endanger life or property                  Comanche Peak Nuclear Power Plant,
                                                                                                                                                                 for leave to intervene must be filed by
                                               or the common defense and security,                      Unit No. 1
                                                                                                                                                                 November 19, 2018.
                                               and is otherwise in the public interest.                 AGENCY:  Nuclear Regulatory                              ADDRESSES: You may submit comments
                                               Therefore, the NRC grants the applicant                  Commission.                                              by any of the following methods:
                                               an exemption from the requirements of                    ACTION: License amendment application;                      • Federal Rulemaking Website: Go to
                                               10 CFR 72.212(a)(2), 72.212(b)(3),                       opportunity to comment, request a                        http://www.regulations.gov and search
                                               72.212(b)(5)(i), 72.212(b)(11), and                      hearing, and petition for leave to                       for Docket ID NRC–2018–0205. Address
                                               72.214 only with regard to meeting TS                    intervene.                                               questions about Docket IDs in
                                               1.2.5 of Attachment A of CoC No. 1004                                                                             regulations.gov to Jennifer Borges;
                                               for DSCs 11–15.                                          SUMMARY:   The U.S. Nuclear Regulatory                   telephone: 301–287–9127; email:
                                                  This exemption is effective upon                      Commission (NRC) is considering                          Jennifer.Borges@nrc.gov. For technical
                                                                                                        issuance of an amendment to Facility                     questions, contact the individual listed
                                               issuance.
                                                                                                        Operating License No. NPF–87, issued                     in the FOR FURTHER INFORMATION
                                                  Dated at Rockville, Maryland, this 13th day           to Vistra Operations Company LLC (the                    CONTACT section of this document.
                                               September 2018.                                          licensee), for operation of the Comanche                   • Mail comments to: May Ma, Office
                                                  For the Nuclear Regulatory Commission.                Peak Nuclear Power Plant (CPNPP),                        of Administration, Mail Stop: TWFN–7–
                                               John McKirgan,                                           Unit No. 1. The proposed exigent                         A60M, U.S. Nuclear Regulatory
                                               Branch Chief, Spent Fuel Licensing Branch,               amendment would revise CPNPP                             Commission, Washington, DC 20555–
daltland on DSKBBV9HB2PROD with NOTICES




                                               Division of Spent Fuel Management, Office                Technical Specification (TS) 3.8.4, ‘‘DC                 0001.
                                               of Nuclear Material Safety and Safeguards.               [Direct Current] Sources—Operating,’’ to                   For additional direction on obtaining
                                               [FR Doc. 2018–20283 Filed 9–17–18; 8:45 am]
                                                                                                        allow the licensee additional time to                    information and submitting comments,
                                                                                                        replace two affected battery cells in the                see ‘‘Obtaining Information and
                                               BILLING CODE 7590–01–P
                                                                                                        safety-related batteries for CPNPP, Unit                 Submitting Comments’’ in the
                                                                                                        No. 1. Specifically, the proposed one-                   SUPPLEMENTARY INFORMATION section of
                                                                                                        time change would add a Required                         this document.


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Document Created: 2018-09-18 01:19:01
Document Modified: 2018-09-18 01:19:01
CategoryRegulatory Information
CollectionFederal Register
sudoc ClassAE 2.7:
GS 4.107:
AE 2.106:
PublisherOffice of the Federal Register, National Archives and Records Administration
SectionNotices
ActionExemption; issuance.
ContactChristian Jacobs, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; telephone: 301-415-6825; email: [email protected]
FR Citation83 FR 47192 

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