Xcel Energy, Monticello Nuclear Generating Plant Independent Spent Fuel Storage Installation, 39069-39076 [2016-14188]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 81, Number 115 (Wednesday, June 15, 2016)]
[Notices]
[Pages 39069-39076]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2016-14188]


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NUCLEAR REGULATORY COMMISSION

[Docket Nos. 72-58 and 50-263; NRC-2016-0115]


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 
September 29, 2015, 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, shall 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 one loaded Standardized NUHOMS[supreg] 61BTH, DSC 16 (DSC 
16), at the Monticello Nuclear Generating Plant (MNGP) Independent 
Spent Fuel Storage Installation (ISFSI).

ADDRESSES: Please refer to Docket ID NRC-2016-0115 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 Web site: Go to https://www.regulations.gov and search for Docket ID NRC-2016-0115. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: Carol.Gallagher@nrc.gov. For technical questions, contact 
the individual listed in the FOR FURTHER INFORMATION CONTACT section of 
this document.
     NRC's Agencywide Documents Access and Management System 
(ADAMS): You may obtain publicly-available documents online in the 
ADAMS Public Documents collection at https://www.nrc.gov/reading-rm/adams.html. To begin the search, select ``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 pdr.resource@nrc.gov. 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: 
Christian.Jacobs@nrc.gov.

SUPPLEMENTARY INFORMATION:

I. Background

    Northern States Power Company-Minnesota, doing business as Xcel 
Energy (Xcel Energy, or the applicant) is the holder of Facility 
Operating License No. DPR-22, which authorizes operation of the 
Monticello Nuclear Generating Plant (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 Monticello Nuclear Generating 
Plant ISFSI. Under the terms of the general license, the applicant 
stores spent fuel at its ISFSI using the Transnuclear, Inc. (TN) 
Standardized NUHOMS[supreg] dry cask storage system Certificate of 
Compliance (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(b)(3) and 10 CFR 72.212(b)(11) 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 the extent necessary for the applicant to 
transfer DSC 16 into a Horizontal Storage Module (HSM). This would 
permit the continued storage of that DSC for the service life of the 
canister. Specifically, the exemption would relieve the applicant from 
meeting TS 1.2.5 of Attachment A of CoC No. 1004, which requires that 
all DSC closure welds, except those subjected to full volumetric 
inspection, shall 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 liquid 
penetrant test acceptance standards shall 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 liquid penetrant test (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. The DSC 16 
remains on the reactor building refueling floor in a transfer cask 
(TC).

[[Page 39070]]

Xcel Energy has performed phased array ultrasonic testing (PAUT) of the 
closure welds, supported by analysis, as an alternate means for 
verifying the weld quality. The PAUT nondestructive examination (NDE) 
consists of testing performed by qualified personnel, using specific 
procedures and equipment shown by performance demonstration to be 
sufficient to detect the range of potential weld defects that could be 
present in the closure welds. The exemption request, if approved, would 
allow the transfer of DSC 16 into an HSM, and would permit the 
continued storage of that DSC for the service life of the canister. 
Xcel Energy plans to request a separate exemption for the remaining 
DSCs (11-15).
    In a letter dated September 29, 2015, as supplemented January 29, 
2016, and March 29, 2016, the applicant requested an exemption from 
certain parts of the following requirements to allow storage of the DSC 
at the MNGP ISFSI:
     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 Sec.  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.
    Upon review, in addition to the requirements from which the 
applicant requested exemption, the NRC staff determined that exemptions 
from the following requirements are also necessary in order to 
authorize the applicant's request and added the following requirements 
to the exemption for the proposed action pursuant to its authority 
under 10 CFR 72.7, ``Specific exemptions'':
     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 this part.
     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 Sec.  72.214.
     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 allow the applicant to transfer DSC 16 into an 
HSM, and would permit the continued storage of that DSC at the MNGP 
ISFSI for the service life of the canister by relieving the applicant 
of the requirement to meet the liquid penetrant test 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, as well as provisions 
determined to be applicable by the NRC staff, 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 liquid penetrant test 
examinations to be performed on the DSCs to verify the acceptability of 
the closure welds, allowing for transfer of DSC 16 into an HSM, and 
would permit the continued storage of that DSC at the MNGP ISFSI for 
the service life of the canister. This exemption only addresses DSC 16, 
for which the PT test was not performed in accordance with the 
examination procedures specified in TS 1.2.5. Xcel Energy performed 
phased array ultrasonic testing to nondestructively examine the welds, 
and prepared structural analyses based on the actual weld quality to 
verify that the welds would perform their desired function over the 
storage term of the DSC. 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 Certificate of Compliance No. 1004. The DSC under consideration for 
exemption was loaded under Certificate of Compliance 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, and 
confinement capability.
    Materials Review for the Requested Exemption: The applicant 
asserted that there is reasonable assurance of safety for the requested 
exemption for the transfer of DSC 16 to the MNGP ISFSI pad. The 
applicant's assertion of reasonable assurance of safety for the 
transfer of DSC 16 is based on the following:
     Repair and verification activities performed on DSC 16;
     PAUT examination and analysis of accessible lid welds on 
DSC 16;
     Short duration and haul distance of the transfer of DSC 
16, and
     The safest location for DSC 16 is in the HSM.
    The applicant asserts that there is a reasonable assurance of 
safety for the requested exemption for DSC 16 (CoC

[[Page 39071]]

No. 1004, Amendment 10) based on the following:
     Integrity of the fuel (cladding) creates a fission product 
barrier;
     The quality of the welding process employed provides 
indication of development of quality welds;
     The advantages of the multi-layer weld technique which 
includes the low probability for flaw propagation, the subsequent 
covering of weld layer surface flaws and the indication of development 
of quality welds;
     Visual inspections performed on the welds met quality 
requirements;
     The DSC backfill and helium leak testing results verify 
confinement barrier integrity;
     The lack of a failure mechanism that adversely affects 
confinement barrier integrity; and
     Margin of safety is available in the welds when assuming 
conservatively large flaws. These margins are demonstrated by two 
different methods: (1) Structural analysis using an analysis-based 
Stress Allowance Reduction Factor and theoretically-bounding full-
circumferential flaws, and (2) a finite element analysis assuming flaw 
distributions conservatively derived from PAUT examination.
    The applicant stated that the PAUT examination and analysis 
provides an objective review of volumetrically-identified flaw 
indications in the accessible DSC 16 Inner Top Cover Plate (ITCP) and 
Outer Top Cover Plate (OTCP) closure welds. The peak strains in the 
welds remain well below the weld material ductility limit when 
subjected to the accident pressure and drop loads. The peak strains 
have a margin of safety of 3.69 and 3.60 for accident pressure and drop 
loads, respectively. Furthermore, it was shown that the strains in the 
welds remain stable at 150 percent of the original design loads for the 
NUHOMS[supreg] 61BTH DSC. The applicant's analysis accounted for the 
identified ITCP and OTCP closure weld flaws and the uncertainties in 
the PAUT examination. The applicant stated that this approach, which is 
consistent with the NRC's Spent Fuel Project Office Interim Staff 
Guidance-15 (ISG-15), conservatively accounts for any additional 
limitations in the efficacy of the PAUT examinations and also accounts 
for the inaccessible area around the vent and siphon block as well as 
the geometric reflectors at the root and near the toe of the closure 
welds.
    The applicant noted that the proposed exemption applies only to DSC 
16 and is supported by the following reports:
    1. Technical Justification for Phased Array Ultrasonic Examination 
of Dry Storage Canister Lid Welds Report No. 54-PQ-114-001, January 30, 
2015 (AREVA, INC., 2015a).
    2. Technical Report of the Demonstration of UT NDE Procedure 54-UT-
114-000 Phased Array Ultrasonic Examination of Dry Storage Canister Lid 
Welds Report No. 51-9234641-001, January 30, 2015 (AREVA, INC., 2015b).
    3. 61BTH ITCP and OTCP Closure Weld Flaw Evaluation, Calculation 
11042-0205 Revision 3 (AREVA, INC., 2016).
    The NRC staff reviewed Technical Justification for Phased Array 
Ultrasonic Examination of Dry Storage Canister Lid Welds Report No. 54-
PQ-114-001, dated January 30, 2015 (AREVA, INC., 2015a). This report 
provides the detailed technical justification for the use of the PAUT 
system to perform the NDE of the OTCP and ITCP closure welds of DSC 16. 
The NRC staff determined that the technical justification report was 
adequate to justify the use of PAUT to examine the ITCP and OTCP 
closure welds because the report included detailed information on the 
PAUT system design, an assessment of examination sensitivity, flaw 
detection, flaw sizing, identification and effects of influential 
parameters, personnel qualification requirements, components to be 
examined, flaws to be detected, and analysis of flaw detection and flaw 
sizing data. In addition, the NRC staff determined that the report also 
described extensive modeling performed to evaluate PAUT array 
configuration, element arrangements, apertures, frequency, focusing, 
and beam angles to develop probes for the inspections of the ITCP and 
OTCP closure welds. The NRC staff also confirmed that the performance 
of the PAUT system was evaluated using laboratory testing of 
representative mockup containing 22 typical welding manufacturing flaws 
that have the potential to exist in field welds. The NRC staff 
determined that the laboratory testing was adequate to verify the 
performance of PAUT systems because the non-blind mockup contained 
representative ITCP and OTCP closure welds with controlled placement of 
intentional flaws positioned in difficult detection locations such as 
in the weld root and weld toe regions and were generally small in size.
    The NRC staff also reviewed ISG-15, which states that closure lid 
welds examined by ultrasonic testing (UT) must use UT acceptance 
criteria of NB-5332 for pre-service examination and be performed in 
conjunction with the PT of the root and final pass. The ISG-15 also 
states that if progressive PT examination is used without a volumetric 
examination, a stress reduction factor of 0.8 is to be imposed on the 
weld design.
    The NRC staff determined that the reduction factor of 0.8 
considered by the applicant in their finite element analysis is 
sufficient to account for weld flaws that potentially were not detected 
by PAUT, visual inspection and the compliant PT inspection of the OTCP 
final weld pass. The NRC staff reached this determination based on the 
demonstrated ability of the PAUT examination to detect weld flaws on 
both the ITCP and OTCP closure welds including the root pass and the 
final pass shown in the technical justification of using PAUT to 
examine the DSC lid closure welds (AREVA, INC., 2015b). The NRC staff 
noted that the PAUT examination results of the OTCP weld are consistent 
with the PT examination of the OTCP closure weld final pass after 
repair and confirmed that no surface breaking flaws are present. Thus, 
the NRC staff determined that analytical evaluation of the DSC 16 OTCP 
and ITCP closure welds using the flaw sizing results obtained by the 
PAUT examination, combined with the discount of the ASME B&PV Code 
specified minimum elongations for the weld material, is an appropriate 
method to determine the acceptability of the DSC inner and outer lid to 
shell closure welds.
    The NRC staff determined that the PAUT procedure (AREVA, INC., 
2016) was acceptable because the procedure was qualified using a blind 
performance demonstration in accordance with ASME B&PV Code Section V, 
Article 14, T-1424(b) Intermediate Rigor (ASME 2004 edition) that 
qualifies the equipment, procedure, and data analysis personnel for the 
detection and dimensioning of welding fabrication flaws. The NRC staff 
determined that PAUT procedures were also acceptable because: (1) 
Personnel conducting the equipment calibration, data acquisition or 
data analyses must be qualified by the American Society for 
Nondestructive Testing (ASNT); (2) the examination area includes the 
accessible area of the ITCP and OTCP closure welds, and (3) specific 
procedures were developed and demonstrated for both flaw detection and 
flaw sizing scans. The NRC staff determined that the examinations were 
appropriate because: (1) They included >99 percent of the OTCP closure 
weld with the exception of two (2) 0.5-inch long sections that were 
identified as limited examination areas as a result of the two 
longitudinal welds in the canister shell; and (2) the entire ITCP

[[Page 39072]]

closure weld with the exception of the part of the weld located around 
the siphon and vent port block resulting in >90 percent coverage of the 
ITCP closure weld (AREVA, INC., 2016). The NRC staff determined that 
the personnel qualifications for equipment calibration, data 
acquisition and data analysis are sufficient because: (1) Data 
Acquisition Operators require direct supervision of American Society 
for Nondestructive Testing (ASNT) UT Level II or Level III staff; (2) 
both Calibration Personnel and Data Analysis Personnel were required to 
be either ASNT UT Level II or Level III certification; and (3) lead 
personnel responsible for training and review of flaw indications were 
required to be ASNT UT Level III qualified. The NRC staff determined 
that the procedures for the flaw detection scans were adequate, 
because: (1) The procedures used the known geometric features of the 
DSC to identify the correct position of the transducer for complete 
coverage of the closure welds to be examined; and (2) the beams are 
swept through a range of angles at specified increments along the scan 
line in order to achieve coverage of the examination volume. The NRC 
staff determined that the flaw sizing scan procedures were adequate 
because: (1) Raster scans were conducted at the higher frequency 
transducer (increased resolution) with a range of beam angles to 
achieve maximum insonification of the flaw; (2) focal laws were 
programmed for a focal depth equal to the reported flaw depth; (3) the 
acquired data was reviewed to verify that signal saturation had not 
occurred or whether rescanning of the area was necessary to obtain a 
response that would allow accurate flaw sizing; and (4) the flaw length 
and flaw height were determined using prescribed signal thresholds. The 
NRC staff determined that the PAUT minimum attributes for flaw 
detection and characterization provided by the applicant were 
acceptable and are commensurate with NRC confirmatory research findings 
involving PAUT examinations of welds (A.A. Diaz, S.L. Crawford, A.D. 
Cinson, and M.T. Anderson, ``Technical Letter Report, An Evaluation of 
Ultrasonic Phased Array Testing for Reactor Piping System Components 
Containing Dissimilar Metal Welds JCN N6398, Task 2A, PNNL-19018,'' 
Richland, WA; Pacific Northwest National Laboratory, November 2009).
    The NRC staff determined that PAUT data analysis methods provided 
by the applicant were adequate because they included specific 
procedures for flaw detection and flaw sizing necessary to locate and 
size flaws in the ITCP and OTCP closure welds using PAUT. The NRC staff 
determined that the applicant demonstrated the accuracy of the PAUT 
flaw detection and flaw sizing procedures using closure welds mockups 
with imbedded flaws. The NRC staff determined that PAUT procedure 
contained sufficient detail to ensure that the examination can be 
repeated with similar results and provides reasonable assurance that 
the examination could detect and size flaw indications found within the 
closure lid weld volumes.
    The NRC staff reviewed Technical Report of the Demonstration of UT 
NDE Procedure 54-UT-114-000 Phased Array Ultrasonic Examination of Dry 
Storage Canister Lid Welds Technical Report Document 51-9234641-001, 
dated January 30, 2015 (AREVA, INC., 2015b). This report summarizes the 
PAUT performance demonstration on a second ITCP and OTCP weld mockup 
specimen known as the blind mockup. The report states the overall task 
objective is to utilize a PAUT technique for detection and 
characterization of fabrication flaws in the closure lid welds of DSCs. 
The developed procedure was evaluated through a blind performance 
demonstration that included the scanning and data analysis of a secured 
(true-state withheld from examiners) OTCP and ITCP closure weld mockup. 
The blind mockup contained a number of controlled welding fabrication 
flaws similar in size and type to the flaws contained in the non-blind 
mockup, but placed in different locations. The technical report of the 
demonstration identified a calculated probability of detection (POD) of 
97 percent with no missed detections (i.e., none of the known imbedded 
flaws in the blind mockup were missed in the performance demonstration) 
and one false call (i.e., one flaw indication reported by an examiner 
in the blind performance demonstration was incorrect and was not an 
actual imbedded flaw). As previously stated, the use of PAUT procedure 
to inspect DSC closure lid welds for this application was developed in 
accordance with ASME B&PV Code Section V, Article 14, T-1424(b), 
Intermediate Rigor (ASME 2004 edition). Intermediate rigor requires 
that a limited performance demonstration be conducted achieving a flaw 
POD of 80 percent and a false call rate of less than 20 percent. The 
NRC staff finds the demonstration of PAUT procedure to be acceptable, 
because the blind performance demonstration results exceed the criteria 
for acceptable performance listed in ASME B&PV Code Section V, Article 
14, T-1471 Intermediate Rigor Detection Test (ASME 2004 edition).
    The NRC staff reviewed Monticello DSC 16 phased array UT 
examination results that were used as an input to the 61BTH ITCP and 
OTCP Closure Weld Flaw Evaluation CALCULATION 11042-0205, Revision 3 
(AREVA, INC., 2016). The NRC staff determined that the examination 
results were acceptable because:
    1. The examination was conducted in accordance with the PAUT 
examination procedure developed in accordance with ASME B&PV Code 
Section V, Article 14, T-1424(b), Intermediate Rigor (ASME 2004 
edition).
    2. Flaws identified were appropriately characterized in terms of 
flaw length and flaw height. The PAUT examination identified the 
location of the flaws with respect to the geometric features of the DSC 
shell, the ITCP and the OTCP, and closure lid welds.
    3. The largest flaw in the OTCP closure weld was characterized as 
having a height of 0.14 inches which is not greater than the thickness 
of one weld bead and less than the OTCP closure weld critical flaw size 
of 0.29 inches.
    4. The largest flaw in the ITCP closure weld was characterized as 
having a height of 0.11 inches which is not greater than the thickness 
of one weld bead and less than the ITCP closure weld critical flaw size 
of 0.15 inches.
    The NRC staff reviewed the preservice examination requirements of 
ASME B&PV Code Section III NB-5280 (ASME 1998 edition with 2000 
addenda). The NRC staff determined that the PAUT examination results 
identified and sized flaws that exceed the acceptance criteria of NB-
5332 (ASME 1998 edition with 2000 addenda), and NB-5332 is an 
acceptable approach under ISG-15. The applicant stated that the flaws 
identified by the PAUT examination were explicitly included in the 
finite element models as design features. Further, all indications 
found through the PAUT exam were, according to the applicant, 
conservatively characterized as planar and evaluated as such. The NRC 
staff determined that the approach taken by the applicant is 
acceptable, because: (1) The PAUT system was capable of identifying and 
sizing the flaws in the ITCP and OTCP welds with the exception of small 
sections of the OTCP closure weld as a result of longitudinal welds in 
the canister shell and the portion of the ITCP closure weld around the 
siphon and vent block; (2) the size of the flaws used in the analysis 
conservatively bounds the size and distributions of flaws identified by

[[Page 39073]]

PAUT; and (3) the applicant applied a reduction factor of 0.8 on the 
ASME B&PV Code specified minimum elongations to the weld material to 
account for flaws that may not have been detected by the PAUT 
examination.
    As a result of the conclusions discussed above, the NRC staff finds 
that there is adequate material performance of the components important 
to safety for DSC 16, loaded under CoC No. 1004, Amendment No. 10, and 
that DSC 16, as addressed in the exemption request, remains in 
compliance with 10 CFR part 72.
    Structural Review for the Requested Exemption: The partial-
penetration welds of the canister OTCP and the ITCP of the Type 1 
NUHOMS[supreg] 61 BTH DSCs were originally evaluated in accordance with 
the ASME B&PV Code Section III, Subsection NB code limits. After the 
weld repair and verification activities on DSC 16, the applicant 
performed a PAUT examination and documented volumetrically-identified 
flaw indications in the welds. In the Materials Review for the 
Requested Exemption, the staff determined that the PAUT examination 
results were appropriate for analytical modeling. The results provided 
a basis for the applicant to model weld flaw size and distribution in 
performing structural evaluation by analysis. The evaluations and 
resulting conclusions to demonstrate the welds structural performance 
is presented below.
    AREVA Calculation No. 11042-0204, Revision 3, ``Allowable Flaw Size 
Evaluation in the Inner Top Cover Weld for DSC # 16,'' used the ASME 
B&PV Code, Section XI, Appendix C flaw evaluation methodology to 
compute the allowable flaw size for governing Load Case TR-9 of an 
internal pressure of 20 psi plus a 25-g inertia loading associated with 
the DSC corner drop. A theoretical subsurface crack or an equivalent 
surface crack residing in the full circumference around the 0.25-inch 
deep ITCP weld in DSC 16 was assumed to be subject to the radial 
tensile membrane force on the weld. For the membrane stress of 17.08 
ksi resulting from multiplying the calculated stress of 13.14 ksi with 
a service factor, SFm, of 1.3 for Service Level D, the 
applicant determined a 0.15-inch wide allowable flaw size. The staff 
reviewed the analysis assumptions and concludes that the flaw size and 
distribution are conservatively modeled in accordance with the ASME 
B&PV Code Section XI flaw evaluation methodology to demonstrate 
sufficient structural performance margins in the welds.
    In Structural Integrity Associates (SIA) Calculation Package No. 
1301415.301, Revision 0, ``Development of an Analysis Based Stress 
Allowable Reduction Factor (SARF), Dry Shielded Canister (DSC) Top 
Closure Weldments,'' the applicant used a finite element analysis (FEA) 
approach to perform generic evaluation of flaw effects on the weld 
stress performance. Three types of flaw geometry, radial, 
circumferential, and laminar flaws for a range of distribution of flaw 
length, depth, and spacing in the DSC ITCP and OTCP were analyzed. 
Following a commonly acceptable FEA practice to simulate flaws with the 
elements of near zero stiffness, the applicant computed the membrane 
and membrane-plus-bending stress intensities in the welds. By comparing 
the results from the FEA models, with and without flaws, for the 
pressure and side drop load cases, a ratio, or SARF, was determined for 
each critical weld section cut of interest. For the OTCP, the applicant 
computed SARFs for 7 flaw configurations each for the individual 
pressure and side drop loading cases. This established a minimum SARF 
of greater than 0.7 for the through-wall circumferential flaws assumed 
to span an arc length of 2.016 inches with a common arc spacing of 
5.184 inches. From the weld quality review documented in the SIA 
report, No. 1301415.405, ``Expectations for Field Closure Welds on the 
AREVA-TN NUHOMS[supreg] 61BTH Type 1 & 2 Transportable Canister for BWR 
Dry Fuel Storage,'' the applicant determined that only the 
circumferential flaws are potentially representative of the weld 
condition of the ITCP. This provided the basis for postulating a 360 
degree, 50 percent intermittently embedded, through-wall 
circumferential flaw with a 0.006 in\2\ cross section area for the FEA. 
This resulted in the calculated SARFs of 0.945 and 0.931 for the 
pressure and side drop cases, respectively. The staff reviewed the 
modeling assumptions and FEA results and concludes that the FEA method 
is suitable for analyzing the stress performance of the weld as a 
continuum with multiple embedded flaws.
    Using the PAUT flaw indication examination results, the applicant 
performed an FEA to determine the weld structural performance margins, 
in accordance with the ASME Section III code limits, for the ITCP and 
OTCP of DSC 16. As noted in AREVA Calculation No. 11042-0205, Revision 
3, ``61BHT ITCP and OTCP Closure Weld Evaluation,'' two full-
circumferential, bounding flaw sets for the OTCP and one for the ITCP 
were used in the simulation of the flaw indications in the FEA models. 
The first set of the two bounding flaws in the OTCP are 0.14 inches and 
0.195 inches each in height while the second set of the three flaws 
range in height from 0.07 inches to 0.16 inches. The single flaw set 
for the ITCP consists of two bounding flaws, a 0.09-inch high flaw 
between the weld metal and the DSC shell and another 0.11-inch high 
inside the ITCP, but at close proximity to the weld metal.
    Using an elastic-perfectly plastic material property model, the 
applicant evaluated the top cover plates-to-shell welds for three 
governing load cases: (1) Internal pressure loading of 32 psi for 
Service Levels A/B; (2) internal pressure loading of 65 psi for Service 
Level D; and (3) side drop loading of 75 g for Service Level D. Given 
that the potential exists for the weld to undergo material yielding, 
the applicant performed a limit analysis, per the ASME B&PV Code, 
Section III, Paragraph NB-3228.1, ``Limit Analysis,'' provisions, for 
the Service Level A/B, normal and off-normal condition load cases. 
Correspondingly, the rules of ASME B&PV Code Section III, Appendix F, 
Paragraph F-1341.3, ``Collapse Load,'' were used for the Service Level 
D, accident condition load cases. The limit analysis, with elastic-
perfectly plastic material model, revealed that the weld would undergo 
unbounded deformation after the material yielding strength is exceeded.
    To address the potential material rupture associated with large 
weld deformation and, hence, high plastic strain concentrations, the 
applicant performed an elastic-plastic analysis to supplement the 
determination of the weld performance margins for DSC 16. This was 
accomplished 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 large-deformation FEA 
models with embedded flaws in the welds. The elastic-plastic analyses 
resulted in the maximum equivalent plastic strains of 5.97 percent and 
6.09 percent for the Service Level D design pressure of 65 psi and side 
drop of 75 g, respectively. The calculated strains are much smaller 
than the ASME B&PV Code specified minimum elongations of SA-240 Type 
304 stainless steel at 40 percent and E308-XX electrode at 35 percent.
    Additionally, for a conservative determination of margins of 
safety, the applicant considered a load factor of 1.5 to evaluate the 
welds subject to a DSC internal pressure of 100 psi (65 x 1.5 = 97.5 
<100 psi) and a side drop of 122.5 g (75 x 1.5 = 122.5 g). The elastic-
plastic

[[Page 39074]]

analyses, per the ASME B&PV Code, Section III, Paragraph NB-3228.3 
Plastic Analysis provisions, resulted in a peak equivalent plastic 
strain of 12.6 percent for both loading cases. On the basis of the weld 
material elongation limit of 28 percent, a reduction of the ASME B&PV 
Code specified weld elongation limit of 35 percent by a factor 0.8 
(0.35 x 0.8 = 0.28), to account for flaws that may not have been 
detected by the PAUT examination, the applicant calculated the margins 
of safety of 3.69 and 3.60 for the internal pressure and side drop 
loading cases, respectively.
    The NRC staff reviewed the FEA modeling assumptions and concludes 
that the elastic-plastic analysis was implemented with appropriate 
loading conditions and materials properties, as described above. The 
analysis results show that the welds would undergo plastic deformation 
for the Service Level D loading associated with canister internal 
pressure and side drop accident conditions. However, no material 
rupture or breach of DSC confinement boundary at the welds is expected 
because of the large margins of safety against the ASME B&PV Code 
specified elongation limits. For this reason, the staff has reasonable 
assurance to conclude that the ITCP and OTCP welds of DSC 16 have 
adequate structural integrity for the normal, off-normal, and accident 
and natural phenomenon conditions. The NRC staff also finds that the 
retrievability of DSC 16 is ensured based on the demonstration of 
adequate structural integrity discussed above.
    The NRC staff finds that the structural function of DSC 16, loaded 
under CoC No. 1004, Amendment No. 10, addressed in the exemption 
request remains in compliance with 10 CFR part 72.
    Thermal Review for the Requested Exemption: The applicant stated 
that even though nonconforming examinations exist, satisfactory 
completion of the required helium leak test conducted on DSC 16 has 
specifically demonstrated the integrity of the primary confinement 
boundary (ITCP and siphon/vent cover plate) welds. These tests 
(conducted per TS 1.2.4a) specifically demonstrate that the primary 
confinement barrier field welds are ``leak tight'' as defined in 
American National Standards Institute (ANSI) N14.5-1997. The licensee 
stated that, in this respect, the helium leak test demonstrates the 
basic integrity of the confinement barrier and the lack of a through-
weld flaw in the field closure welds that would lead to a loss of 
cavity helium in DSC 16. The licensee 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 DSC 16) which assures the helium atmosphere in the DSC 16 cavity is 
maintained in order to support heat transfer.
    The NRC staff reviewed the licensee's exemption request and also 
evaluated its effect on the DSC 16 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 DSC 16, which assures 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. Therefore, based on the NRC 
staff's review of the licensee's evaluation and technical 
justification, the NRC staff finds the exemption request acceptable by 
virtue of the demonstrable structural integrity of the ITCP and OTCP.
    The NRC staff finds that the thermal function of DSC 16, 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 DSC 16 presented in the Monticello 
exemption request. The NRC staff finds that DSC 16 is not affected by 
the nonconforming PT examinations because storage of DSC 16 on the MNGP 
ISFSI will not significantly alter the assumptions of the criticality 
safety and radiation protection analysis of the 61BTH DSC. The interior 
of DSC 16 will continue to prevent water in-leakage, which means that 
the system will remain subcritical under all conditions. The 
nonconforming PT examinations do not affect the radiation source term 
of the spent fuel contents, or the configuration 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 DSC 16, 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 DSC 16 loaded at the 
MNGP met the confinement-related requirements described in 10 CFR part 
72.
    As described in the licensee's ``Exemption Request for 
Nonconforming Dry Shielded Canister Dye Penetrant Examinations'' 
(Enclosure 1 of the September 29, 2015, submittal), certain elements of 
the DSC 16 closure weld PT examinations did not comply with examination 
procedures. To support the exemption request, the licensee 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 1E-7 
cc/sec (i.e., ``leaktight'' as defined by ANSI N14.5). The applicant 
noted that failure to comply with the PT examination procedures would 
not change the general integrity of these DSC closure welds. 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 and, therefore, the demonstration of the closure 
confinement integrity, as defined by the licensing basis, is 
unaffected. In addition, in the Structural Review for the Requested 
Exemption and Materials Review for the Requested Exemption evaluations 
described previously, staff evaluated the applicant's repair and 
verification activities and the PAUT examinations and analyses 
associated with DSC 16 and concluded DSC 16 meets the requirements of 
10 CFR part 72.
    As discussed above, because the PT examinations did not affect DSC 
16's helium leak test results, the NRC staff finds that the confinement 
function of DSC 16, loaded under CoC No. 1004, Amendment No. 10, 
remains in compliance with 10 CFR part 72.
    Review of Common Defense and Security: The NRC staff considered the 
potential impacts of granting the exemption on the common defense and 
security. The requested exemption is not related to any security or 
common defense aspect of the MNGP ISFSI, therefore granting the 
exemption would not result in any potential impacts to common defense 
and security.
    Based on its review, the NRC staff has reasonable assurance that 
the storage system will continue meet the thermal, structural, 
criticality, retrievability and radiation protection requirements of 10 
CFR part 72 and, therefore, will not endanger life or property. The NRC 
staff

[[Page 39075]]

also finds that there is no threat to the common defense and security.
    Therefore, the NRC staff concludes that the exemption to relieve 
the applicant from meeting TS 1.2.5 of Attachment A of CoC No. 1004, 
Amendment No. 10, which requires that liquid penetrant test 
examinations be performed on DSCs to verify the acceptability of the 
closure welds, allowing for transfer DSC 16 into an HSM, and would 
permit the continued storage of that DSC for the service life of the 
canister at the MNGP ISFSI will not endanger life or property or the 
common defense and security.

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) DSC 16 would have to be opened and unloaded, and 
the contents loaded in a new DSC, and that DSC welded and tested, or 
(2) the OTCP would need to be machined off, and the ITCP weld machined 
down to the root weld; and the 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). If there were 
no such damage, the DSC would need to be re-welded and inspected. Both 
options would entail a higher risk of a cask handling accidents, 
additional personnel exposure, and greater cost to the applicant. Both 
options would also generate additional radioactive contaminated 
material (including the unloaded DSC for option 1) and waste from 
operations, because the lid would have to be removed in either case, 
which would generate cuttings from removing the weld material that 
could require disposal as contaminated material.
    The proposed exemption to allow transfer of DSC 16 into an HSM, and 
permit the continued storage of that DSC for the service life of the 
canister at the MNGP ISFSI, is consistent with NRC's mission to protect 
public health and safety. Approving the requested exemption produces 
less of an opportunity for a release of radioactive material than the 
alternatives to the proposed action because there will be no operations 
involving opening the DSCs which confine the spent nuclear fuel. 
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 Technical 
Specification (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 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.
------------------------------------------------------------------------
Monticello Nuclear Generating Plant      ML15275A023
 Exemption Request for Nonconforming     ML15275A024
 Dry Shielded Canister Dye Penetrant     ML15275A025
 Examinations, September 29, 2015.
Monticello Nuclear Generating Plant      ML16035A214
 Exemption Request for Nonconforming     ML16049A081
 Dry Shielded Canister Dye Penetrant     ML16049A094
 Examinations, Supplemental
 Information, January 29, 2016.
Monticello Nuclear Generating Plant      ML16091A228
 Exemption Request for Nonconforming     ML16097A460
 Dry Shielded Canister Dye Penetrant
 Examinations, Supplemental Information
 to Respond to the Second Request for
 Additional Information, March 29, 2016.
Interim Staff Guidance No. 15, Rev. 0,   ML010100170
 Materials Evaluation, January 10, 2001.
Technical Justification for Phased       ML16035A185
 Array Ultrasonic Examination of Dry     ML16035A186
 Storage Canister Lid Welds Report No.   ML16049A094
 54-PQ-114-001, January 30, 2015.
Technical Report of the Demonstration    ML16035A184
 of UT NDE Procedure 54-UT-114-000
 Phased Array Ultrasonic Examination of
 Dry Storage Canister Lid Welds Report
 No. 51-9234641-001, January 30, 2015.
61BTH ITCP and OTCP closure Weld Flaw    ML16097A460
 Evaluation, Calculation 11042-0205,
 Revision 3, March 21, 2016.
Technical Letter Report, An Evaluation   ML093570315
 of Ultrasonic Phased Array Testing for
 Reactor Piping System Components
 Containing Dissimilar Metal Welds JCN
 N6398, Task 2A, PNNL-19018,''
 Richland, WA; Pacific Northwest
 National Laboratory, November 2009.

[[Page 39076]]

 
AREVA Calculation No. 11042-0204,        ML15275A024
 Revision 3, Allowable Flaw Size
 Evaluation in the Inner Top Cover Weld
 for DSC #16, September 29, 2015.
Structural Integrity Associates          ML15275A025
 Calculation Package No. 1301415.301,
 Revision 0, Development of an Analysis
 Based Stress Allowable Reduction
 Factor (SARF), Dry Shielded Canister
 (DSC) Top Closure Weldments, October
 2014.
Structural Integrity Associates report,  ML14309A194
 No. 1301415.405, Expectations for
 Field Closure Welds on the AREVA-TN
 NUHOMS[supreg] 61BTH Type 1 & 2
 Transportable Canister for BWR Dry
 Fuel Storage, November 3, 2014.
------------------------------------------------------------------------

IV. 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), 72.214, and 72.212(b)(11), only with 
regard to meeting Technical Specification (TS) 1.2.5 of Attachment A of 
CoC No. 1004 for DSC 16.
    This exemption is effective upon issuance.

    Dated at Rockville, Maryland, this 8th day June, 2016.

For the Nuclear Regulatory Commission.

Bernie White,
Acting Branch Chief, Spent Fuel Licensing Branch, Division of Spent 
Fuel Management, Office of Nuclear Material Safety and Safeguards.
[FR Doc. 2016-14188 Filed 6-14-16; 8:45 am]
 BILLING CODE 7590-01-P