[Federal Register Volume 83, Number 181 (Tuesday, September 18, 2018)]
[Notices]
[Pages 47192-47203]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[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