[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 http://www.regulations.gov and search for Docket ID NRC-2016-0115. Address
questions about NRC dockets to Carol Gallagher; telephone: 301-415-
3463; email: [email protected]. 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 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]. 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 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