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ANAC ffffl INTERNATIONAL
Atlanta Corporate Headquarters: 3950 East Jones Bridge Road, Norcross, Georgia 30092 USA Phone 770-447-1144, Fax 770-447-1797, www.nacintl.com
June 2018
Revision 18B
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Enclosure 1 to ED20180063 Page 1 of 1
Enclosure 1
RAJ Responses for
NAC-STC SAR, Revision 18B
June 2018
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NACINTERNATIONAL
NON-PROPRIETARY RESPONSES TO THE
UNITED ST ATES NUCLEAR REGULATORY COMMISSION
REQUEST FOR ADDITIONAL INFORMATION
February 2018
NAC-STC Docket No: 71-9235
CoC No: 9235
FOR REVIEW OF THE CERTIFICATE OF COMPLIANCE NO. 9235, STC TRANSPORATION PACKAGE
(CoC NO. 9235 DOCKET NO. 71-9235)
June 2018
Page 1 of 18
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TABLE OF CONTENTS
NAC-STC Docket No: 71-9235
Coe No: 9235
GENERAL INFORMATION ....................................................................................................................... 3
MATERIALS EVALUATION ..................................................................................................................... 4
THERMAL EVALUATION ........................................................................................................................ 8
ACCEPTANCE AND MAINTENANCE TESTS EVALUATION ........................................................... 10
Page 2 of 18
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NAC INTERNATIONAL RESPONSE TO
REQUEST FOR ADDITIONAL INFORMATION
GENERAL INFORMATION
NAC-STC Docket No: 71-9235
CoC No: 9235
1.1 Clarify callout to Note 2 in Drawing No. 423-870 Sheet 1 of I Revision 7 in Zone El.
Note 2 was deleted from the drawing but the callout "SEE NOTE 2" remains in the
revised drawing.
This information is needed to determine comp! iance with Title IO of the Code of Federal Regulations (10 CFR) 71.31 ( c ).
NAC International Response to Acceptance and Maintenance Tests Evaluation RAI 1.1:
Reference to Note 2 in drawing zone El has been removed from the field of the drawing .
Page 3 of 18
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NAC INTERNATIONAL RESPONSE
TO REQUEST FOR ADDITIONAL INFORMATION
MATERIALS EVALUATION
NAC-STC Docket No: 71-9235
CoC No: 9235
2.1 Clarify the operating temperatures for ASTM International (ASTM) A276 Type 304 SS
which was added to the bill of materials for Item #2 in Drawing No. 423-859, Revision 1, Sheet 1 of 1. Note that ASTM A276 304 SS is not included in American Society of
Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code Section II Part
D. Provide mechanical prope11ies over the expected range of operating temperatures if
the mechanical properties of this component are necessary at elevated temperatures.
This information is needed to determine compliance with IO CFR 71.31 (c) and 10 CFR 71.33(a)(5).
NAC International Response to Acceptance and Maintenance Tests Evaluation RAI 2.1:
The use of ASTM A276 304 SS for the retaining rod washer material of the balsawood
impact limiter (Item 2, Drawing No. 423-859 Revision I) is consistent with retaining rod
washer material approved for the redwood impact I imiter (Item 9, Drawing No. 423-811
referenced on the package assembly via Drawing No. 423-900). Both washers are used
for the same function of attaching and retaining the respective impact limiter. Therefore,
the addition of the material option for the balsawood impact limiter configuration is consistent with the current approved system. As detailed in Chapter 9 of the STC SAR,
the STC transport system is designed in accordance with "ASME Boiler and Pressure Vessel Code," The American Society of Mechanical Engineers, 1989 and 1992, with
Addenda, for the STC cask directly loaded fuel configurations
Page 4 of 18
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NAC INTERNATIONAL RESPONSE TO
REQUEST FOR ADDITIONAL INFORMATION
MATERIALS EVALUATION
NAC-STC Docket No: 71-9235
CoCNo: 9235
2.2 Update the specification for the forged stainless steel (SS). Table 2.3.2-2 identifies Type
304 SS under specification SA-336. The austenitic SS that were formerly listed under
SA-336 have been moved to SA-965 in 2008. SA-336 Type 304 SS is called out in
Drawing Nos. 423-802 sheet 1 of7 and 423-804 Sheet 1 of 3.
This information is needed to determine compliance with 10 CFR 71.31(c) and 10 CFR
71.33(a)(5).
NAC International Response to Acceptance and Maintenance Tests Evaluation RAJ 2.2:
As detailed in Chapter 9 of the STC SAR, the STC transport system is designed in
accordance with "ASME Boiler and Pressure Vessel Code," The American Society of
Mechanical Engineers, 1989 and 1992, with Addenda, for the STC cask directly loaded
fuel configurations. Therefore, the ASME SA-336 Type 304 SS material reference is
consistent with the design basis code year and no change is necessary. Note that later
versions of the ASME SA-336 material specification (post 2001 with the 2003 addenda)
provides direction to SA-965 for stainless steel forgings .
Page 5 of 18
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NAC INTERNATIONAL RESPONSE TO
REQUEST FOR ADDITIONAL INFORMATION
MATERIALS EVALUATION
NAC-STC Docket No: 71-9235
CoC No: 9235
2.3 Revise Tables 2.3.2-2 and 2.3.2-3 footnotes to identify the correct sources of information
from the ASME B&PV code. Except for fatigue design tables and figures, the mechanical property tables in Section Ill Appendix I have been moved to Section II Part D in the 1992 addenda.
This information is needed to determine compliance with 10 CFR 71.31 ( c) and IO CFR 71.33(a)(5).
NAC International Response to Acceptance and Maintenance Tests Evaluation RAJ 2.3:
The footnotes of the tables in Section 2.3.2 will be revised to note the corresponding tables from ASME B&PV Code Section II, Part D as follows:
Table 2.3.2-1 Mechanical Properties of SA 240, Type 304 Stainless Steel
1 "ASME Boiler and Pressure Vessel Code," Section 11, Part D, Table U. 2 "ASME Boiler and Pressure Vessel Code," Section Il, Part D, Table Y-1. 3 "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table 2A. 4 "ASME Boiler and Pressure Vessel Code," Section 11, Paii D, Table TM-I. s "ASME Boiler and Pressure Vessel Code," Section lIJ, Appendix I, Table 1-9.1. 6"ASME Boiler and Pressure Vessel Code," Section II, Part D, Table TE-1. 1 "ASME Boiler and Pressure Vessel Code," Section 11, Part D, Table NF-1.
Table 2.3.2-2 Mechanical Properties of SA 336, Type 304 Stainless Steel
1 "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table U. 2 "ASME Boiler and Pressure Vessel Code," Section 11, Part D, Table Y-1. 3 "ASME Boiler and Pressure Vessel Code," Section ll, Part D, Table 2A.
4 "ASME Boiler and Pressure Vessel Code," Section JI, Part D, Table TM-1.
s "ASME Boiler and Pressure Vessel Code," Section III, Appendix I, Table I-9.1.
6 "ASME Boiler and Pressure Vessel Code," Section II, Pa1i D, Table TE-1. 1 "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table NF-1. s "Nuclear Materials Handbook," Volume 1, Design Data, Property Code 3304 .
Page 6 of 18
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RAI 2.3 Continued
NAC-STC Docket No: 71-9235
CoCNo: 9235
Table 2.3.2-3 Mechanical Propetiies of Type XM-19 Stainless Steel
1 "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table U.
2 "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table Y-1.
J "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table 2A.
4 "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table TM-1.
s "ASME Boiler and Pressure Vessel Code," Section III, Appendix I, Table I-9.1.
6 "ASME Boiler and Pressure Vessel Code," Section JI, Part D, Table TE-I.
7 "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table NF-1.
s "Nuclear Materials Handbook," Volume 1, Design Data, Property Code 3304.
9 SA-182, FXM-19 stainless steel may be substituted for SA-240 XM-19 stainless steel provided that
the SA-182 material yield and ultimate strengths are equal to or greater than those of the SA-240
material. The SA-182 forging material and the SA-240 plate material are both XM-19 austenitic
stainless steels. Austenitic stainless steels do not experience a ductile-to-brittle transition for the
range of temperatures considered in this Safety Analysis Report. Therefore, fracture toughness is
not a concern .
Page 7 of 18
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NAC INTERNATIONAL RESPONSE TO
REQUEST FOR ADDITIONAL INFORMATION
THERMAL EVALUATION
NAC-STC Docket No: 71-9235
CoC No: 9235
3 .1 Provide justification and, if necessary, calculations to show how the thermal analysis in
Section 3.4.1.1.1.3, "Radial Neutron Shield," of the application remains bounding for
normal conditions of transport and hypothetical accident conditions calculated
temperatures considering the change to note 7 on Drawing No. 423-802, Sheet 1.
No justification and calculations were provided to show how the normal conditions of
transport and hypothetical accident conditions calculated temperatures remain bounding
considering the change, "Alternate pre-bonded thickness of 8mm to 10mm for 304 SS
and 6mm to 8mm for copper plates may be used.," to note 7 on Drawing No. 423 802,
Sheet 1.
This information is necessary to demonstrate compliance with 10 CFR 71.71 and 10 CFR
71.73 .
NAC International Response to Acceptance and Maintenance Tests Evaluation RAI 3.1:
The effective thermal conductivity calculation for the radial neutron shield presented in SAR Section 3 .4.1.1.1.3 is based on the 8 mm stainless steel plate and 6 mm copper plate. The change in note 7 of Drawing No. 423-802, Revision 23 allows the thickness for the stainless steel plate and copper plate to be increased to 10 mm and 8 mm respectively. The thermal analysis for normal conditions of transport for the directly loaded STC configuration is performed using the thermal models described in SAR Section 3.4.1.1 and the maximum temperatures are provided in Section 3.4.2.1 (Tables 3.4-1, 3.4-2 and 3 .4-3). The increased effective thermal conductivity for the radial neutron shield provides a slightly more effective path for heat rejection from the cask shell to the ambient. This results in a slight reduction of the maximum temperatures presented in Section 3.4.2.1. Therefore, the temperatures presented in Section 3.4.2.1 remain bounding and no further analysis is required.
For the hypothetical accident, the thermal evaluation for the directly loaded STC
configuration uses the thermal model described in SAR Section 3 .5 .1.1.1 and the maximum
component temperatures are presented in Section 3 .5 .3 (Table 3 .5-1 ). The radial neutron
shield property used in the thermal model is discussed in Section 3.5.1.1.3. Note that the NS-4-FR in the neutron shield is conservatively considered to be present for the entire
Page 8 of 18
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RAI 3.1 Continued
NAC-STC Docket No: 71-9235
CoC No: 9235
30-minute fire condition to maximize the heat input to the model. At the end of the fire transient, the neutron shield is considered to be voided ofNS-4-FR, leaving only the
stainless steel/copper fins and stainless steel shell. The effective conductivity of the
neutron shield is calculated using the same method described in SAR Section 3.4.1.1.1.3 with the NS-4-FR material substituted by air. Note that the temperature results for steady state thermal analysis for the normal condition of transport are used as the initial condition for the fire transient analysis. As previously discussed, the temperatures for the normal condition will be reduced due to higher effective conductivity of the radial neutron shield. Due to the short duration of the fire (30 minutes), cask shells and contents peak temperatures occur after 30 minute fire condition. The heat rejection after
the fire is slightly enhanced. The drawing change to allow thicker stainless steel and copper plates will have an insignificant impact on the maximum component temperatures
for the hypothetical fire accident as presented in SAR Table 3 .5-1.
To further evaluate the effect of the drawing change allowing thicker stainless steel and coper plates in the radial neutron shield, a sensitivity analysis is performed using a threedimensional thermal model corresponding to the limiting configuration and heat load of the NAC-STC transport cask, i.e. the STC-HBU configuration as approved by CoC No . 9235, Revision No. 15, December 20, 2016. The three-dimensional model from the main body ofNAC Calculation No. 423-3000 Rev. 5 is used to perform a steady state analysis for the normal condition of transport, as well as a transient analysis for the fire accident. The model is a 180° half-symmetry full length model for the cask containing the loaded
basket. The governing Case H3 (see Calculation No. 423-3000) with the design basis heat load of 24 kW for the STC-HBU is used. For the sensitivity study, the only change
in the thermal model is the effective conductivity of the radial neutron shield, which were re-calculated using same method as described in SAR Section 3 .4.1.1.1.3 considering the increased thickness of stainless steel and copper plates. The analysis results indicate that the maximum fuel temperature decreased by 4°F (from 638°F to 634°F) for the normal condition of transport. The maximum fuel temperature for the fire accident remained approximately the same (decreased only 1 °F, from 698°F to 697°F). See Appendix AA ofNAC Calculation No. 423-3000 Rev. 6 for details of the steady state and transient analysis.
It is concluded that the change in note 7 on Drawing No. 423-802, Revision 23, Sheet 1
has an insignificant effect on the thermal performance of NAC-STC cask. The analysis results presented in SAR Section 3.4.1.2 for normal condition of transport and Section
3.5.3 for hypothetic fire accident remain bounding and no revision is required .
Page 9 of 18
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NAC INTERNATIONAL RESPONSE TO
REQUEST FOR ADDITIONAL INFORMATION
ACCEPTANCE AND MAINTENANCE TESTS EVALUATION
NAC-STC Docket No: 7 l-9235
CoC No: 9235
8.1 Provide a justification for the position implied in the SAR that a visual inspection of the
radial neutron shield shell, described in Sections 8.2.6 and 8.2.7 of the application, would
provide the necessary information to detect deterioration of the heat transfer properties of
the package. In addition, provide specific acceptance criteria for visual inspections of an
NAC-STC packaging, include a demonstration that these criteria are adequate to verify
the thermal performance of the packaging.
Sections 8.2.6, "Post-fabrication Thermal Test," and 8.2.7, "Miscellaneous," of the
application do not provide a justification that the visual inspection of the radial neutron
shield shell will be able to detect conditions that might lead to the deterioration of the
heat transfer properties of the package, especially any deterioration of the fins that are
internal to the radial neutron shield shell and not visible by inspecting the radial neutron
shield shell. A specific acceptance criteria for the visual inspection, e.g. a more specific
acceptance criteria than described in Section 8.2. 7 of the application, " ... any crack, gauge
(assume "gouge" is meant), or gross deformation that could indicate damage of the heat
transfer fins ... ," was not provided. Also, it has not been demonstrated that visual
inspection is an effective method, with either general or specific acceptance criteria, in
order to verify that the thermal performance of the package has not deteriorated.
Table 3.8-4, "Maximum Component Temperatures - Normal Transport Conditions,
Maximum Decay Heat, Maximum Ambient Temperature, among Three Configurations -
the STC-HBU," of the SAR dated April 11, 2017 (ADAMS Accession No.
ML 171 l 6A075) shows that the radial neutron shield maximum temperature is 295°F,
which is close to the upper temperature limit for the radial neutron shield of 300°F as
indicated in the SAR (Section 3.8.3.2, "Safe Operating Range").
The heat transfer capabilities of the fabricated packaging will impact the overall thermal
performance of the assembled package and in order for components, like the radial
neutron shield, to remain below the allowable temperature limit, the package must be in a
condition commensurate with its design at all times; therefore, it is important for all
inspections to be able to determine if there could be any potential degradation in thermal
performance.
This information is necessary to demonstrate compliance with 10 CFR 71.71 .
Page 10 of 18
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RAI 8-1 Continued
NAC-STC Docket No: 71-9235
CoC No: 9235
NAC International Response to Acceptance and Maintenance Tests Evaluation RAI 8.1:
In understanding the basis for selecting a visual inspection of the neutron shield shell as
the approach to determine continuing compliance with the casks thermal heat rejection
design, it is important to understand the mechanical attachment of the thermal fins to the
cask body and the neutron shield shell, as well as the nature of the materials used and the
extent of cask normal conditions with which cask operations are performed. The
robustness of the design would preclude any degradation of the weldments due to normal
cask operations. Any issues associated with the adequacy of the fabrication would be
discovered on the first article thermal inspection.
The fin is an explosive-bonded, bi-metallic ( copper & 304 stainless steel), component
which is attached to the cask body outer shell (304 stainless steel) using a full penetration
groove weld with 1/8" fillet reinforcement. The weld is PT examined per Section V,
Article 6, with acceptance criteria per Section III, NF-5350.
The neutron shield shell is then attached to the using one of two approved methods, full
penetration groove welds or a single full penetration double-bevel weld. Again, the weld
is penetrant examined per Section V, Article 6, with acceptance criteria per Section III,
NF-5350.
Thermal cycling would not impose significant stresses at the connections of these
members due to their being similar materials. Normal operations will not fail these
connections as there is no loading of the connections described. The instance of an impact capable of imposing loads resulting in deformation or weld failure would be clearly visible by means of a general visual inspection of the neutron shield .
Page 11 of 18
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NAC INTERNATIONAL RESPONSE TO
REQUEST FOR ADDITIONAL INFORMATION
ACCEPTANCE AND MAINTENANCE TESTS EVALUATION
NAC-STC Docket No: 71-9235
CoC No: 9235
8.2 Provide a description (in Section 8.2.6 of the application) of how the NAC-STC package
will be monitored during handling and transportation operations for the normal
conditions of transport described in 10 CFR 71.7l(c).
Section 8.2.6, "Post-fabrication Thermal Test," of the application does not demonstrate
how the package will be monitored during handling and transp011ation operations to
ensure compliance with each of the requirements in 10 CFR Part 71.71 (c). For example,
the heat condition (an ambient temperature of 100°F in still air with solar insolation)
could potentially be exceeded during transportation operations, yet without monitoring,
this would not be detected. In addition, monitoring of the 10 CFR 71.7l(b) initial
conditions may be necessary (see RAI 3, below).
This information is necessary to demonstrate compliance with 10 CFR 71.7l(c) .
NAC International Response to Acceptance and Maintenance Tests Evaluation RAI 8.2:
There is no regulatory requirement to monitor the package to ensure compliance with 10
CFR Part 71.71 (c). NAC recognizes the importance of maintaining the package within the requirements of IO CFR Pai1 71. 71 ( c) during transport, however, it was not the intent
to state that a post-fabrication thermal test would be performed after exceeding the each
of the requirements of 10 CFR Part 71.7l(c). Therefore, NAC has revised the text in
Section 8.2.6 to read as follows:
"However, a post-fabrication thermal test shall be performed on an operational NAC
STC packaging if, during handling or transport operations, the packaging experiences an
adverse event such as fire, drops or impacts that result in obvious damage to the neutron shield."
Page 12 of 18
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NAC INTERNATIONAL RESPONSE TO
REQUEST FOR ADDITIONAL INFORMATION
ACCEPTANCE AND MAINTENANCE TESTS EVALUATION
NAC-STC Docket No: 71-9235
CoC No: 9235
8.3 Address the 10 CFR 71.7l(b) initial conditions in Section .2.6 of the application and
describe action(s) taken if those conditions are exceeded during transport operations.
Section 8.2.6, "Post-fabrication Thermal Test," of the application describes that a thermal
test will be performed on an operational NAC-STC packaging if the conditions or tests of
10 CFR 71.71 are exceeded during transportation operations. Section 8.2.6 of the
application does not describe if any action is taken if the initial conditions in IO CFR
71.71 (b) are exceeded (see item 8-2, above to address monitoring of the 10 CFR 71.71 (b)
initial conditions during handling and transport operations), or describe why not taking any action(s) is justified.
This information is necessary to demonstrate compliance with 10 CFR 71.71(b) .
NAC International Response to Acceptance and Maintenance Tests Evaluation RAJ 8.3:
There is no regulatory requirement to monitor the package to ensure compliance with I 0
CFR Part 71.71 (c). NAC recognizes the importance of maintaining the package within
the requirements of 10 CFR Pai1 71.71 (c) during transport, however, it was not the intent
to state that a post-fabrication thermal test would be performed after exceeding the each
of the requirements of 10 CFR Part 71.7l(c). Therefore, NAC has revised the text in Section 8.2.6 to read as follows:
"However, a post-fabrication thermal test shall be performed on an operational NAC
STC packaging if, during handling or transport operations, the packaging experiences an
adverse event such as fire, drops or impacts that result in obvious damage to the neutron shield."
Page 13 of 18
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NAC INTERNATIONAL RESPONSE TO
REQUEST FOR ADDITIONAL INFORMATION
ACCEPTANCE AND MAINTENANCE TESTS EVALUATION
NAC-STC Docket No: 71-9235
CoC No: 9235
8.4 Clarify the acceptance criteria for the outer closure lid and the outer bottom plate. The
applicant stated: "The outer closure lid and the outer bottom plate will be UT [ultrasonic
tested] to demonstrate their soundness as gamma shielding utilizing ASME Section V,
Article 23, acceptance criteria. Plate shall be accepted per NB-2530 and forgings will be
accepted per NB-2540. Each ASME specification provides requirements for testing
equipment, test method, acceptance criteria, material traceability, and supporting
documentation. SAR Section 8.1.5 .1 has been revised to reflect these requirements."
ASME B&PV Section V Article 23 is not referenced in Section Ill NB-2530 or NB-2540.
It is only referenced in NB-2585 with respect to the examination of bolts. Note that
NB 2532.1 references SA-578 which is included in Section V Article 23. NB-2540
references Section V Article 5. NB-2540 does not reference ASME B&PV Section V
Article 23 .
It appears that the appropriate examination and acceptance criteria would be as follows:
The outer closure lid will be UT examined in accordance with ASME B&PV NB-2542.1
and the acceptance standards of Section NB-2542.2. The outer bottom plate shall be
examined in accordance with NB-2532.1 with the acceptance standards of NB-2532.1 (b).
This information is needed to determine compliance with IO CFR 71.31 ( c) and 10 CFR
71.33(a)(5) and 10 CFR 71.5l(a)(2).
NAC International Response to Acceptance and Maintenance Tests Evaluation RAJ 8.4:
NAC concurs with the reviewer's observation and acknowledges the Section V, Article
23 and Article 5 disconnect. NAC has revised Section 8.1.5.1, paragraph 2, to read as
follows:
"A gamma scan test is not required for the cask inner closure lid, cask outer closure lid,
cask inner bottom forging, cask outer bottom forging, or cask outer bottom plate. These
components shall be ultrasonic tested to demonstrate their soundness as gamma shielding.
Ultrasonic testing shall be pe,formed per ASME B&PV NB-2542. 1 using the acceptance standards of Section NB-2542.2for forgings andASME B&PV NB-2532.1 using the acceptance standards of NB- 2532.l(b)for plates."
Page 14 of 18
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NAC INTERNATIONAL RESPONSE TO
REQUEST FOR ADDITIONAL INFORMATION
ACCEPTANCE AND MAINTENANCE TESTS EVALUATION
NAC-STC Docket No: 71-9235
Coe No: 9235
8.5 Clarify the acceptance criteria for UT of the package outer bottom plate and package
inner bottom forging welds to the package outer bottom forging. The applicant stated:
"The package outer bottom forging will be UT to demonstrate its soundness as a gamma
shield utilizing ASME B&PV Section V, Article 23, acceptance criteria. The forging will
be accepted per NB-2540." The ASME B&PV specification provides requirements for
testing equipment, test method, acceptance criteria, material traceability, and supporting
documentation. SAR Section 8.1.5 .1 has been revised to add the package outer bottom
forging to the UT requirements."
NB-2540 references ASME B&PV Section V, Article 5. NB-2540 does not reference ASME B&PV Section V, Article 23.
This information is needed to determine compliance with 10 CFR 71.31 ( c ) .
NAC International Response to Acceptance and Maintenance Tests Evaluation RA! 8.5:
As indicated in the response to Item 8.4, NAC concurs with the reviewer's observation
and acknowledges the Section V, Article 23 and Article 5 disconnect. NAC has revised 8.1.5.1, paragraph 2, to read as follows:
"A gamma scan test is not required for the cask inner closure lid, cask outer closure lid,
cask inner bottom forging, cask outer bottom forging, or cask outer bottom plate. These components shall be ultrasonic tested to demonstrate their soundness as gamma shielding.
Ultrasonic testing shall be performed per ASME B&PV NB-2542.1 using the acceptance standards o_[Section NB-2542.2for forgings and ASME B&PV NB-2532.1 using the acceptance standards of NB- 2532.1 (b) for plates."
Page 15 of 18
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NAC INTERNATIONAL RESPONSE TO
REQUEST FOR ADDITIONAL INFORMATION
ACCEPTANCE AND MAINTENANCE TESTS EVALUATION
8.6 Provide additional clarification for the alternate lead pour procedures.
NAC-STC Docket No: 71-9235
CoC No: 9235
Section 8.4.2.2 states that during the lead pour the bottom end of the filler-tube is kept
below the surface of the molten lead to preclude the formation of voids in the lead.
Clarify whether the same practice is necessary for the alternate procedure.
Section 8.4.3.2 states that the body weldment will be heated in a steady, uniform, and
controlled manner. Provide the allowable heating rates.
Section 8.4.3.2 states that the temperature of the entire body weldment is maintained
between 640°F (338°C) and 740°F (393°C) throughout the lead pour operations,
approximately. Provide clarification on what "approximately" is referring to in this
context.
Section 8.4.3.3 states that the cooldown rate is held steady, uniform and controlled
manner. Provide maximum cooldown rate and the maximum allowable temperature differential between the inner and outer shell.
This information is needed to determine compliance with IO CFR 71.31 ( c) and
71.33(a)(5).
NAC lnternational Response to Acceptance and Maintenance Tests Evaluation RA] 8.6:
The response to this RAJ was provided on March 6111 in submittal 18A .
Page 16 of 18
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NAC INTERNATIONAL RESPONSE TO
REQUEST FOR ADDITIONAL INFORMATION
ACCEPTANCE AND MAINTENANCE TESTS EVALUATION
NAC-STC Docket No: 71-9235
CoC No: 9235
8.7 Clarify acceptable testing methods for the transport impact limiter SS shell. Section
8.1.4.3 states that a leak test of the shell welds shall be performed to verify weld integrity.
Three acceptable test methods are included but only the first two are actually leak tests.
The third method listed is penetrant testing which is a non-destructive test method. If
penetrant testing is allowed in lieu of an actual leak test, provide the penetrant testing
acceptance criteria and explain why the penetrant testing is a suitable method in lieu of a
leak test.
This information is needed to determine compliance with IO CFR 71.31 ( c ).
NAC International Response to Acceptance and Maintenance Tests Evaluation RAJ 8.7:
It is intended that the impact limiter shells are not to be qualified as "containment or
pressure retaining vessels", but sealed against environmental effects during transport,
specifically ingress of moisture. The implementation of the leakage test allows a single
point evaluation of the shell integrity. Implementation of a penetrant test for pin-holes,
cracks and/or porosity will also provide a similar level of confidence in the shells
environmental integrity.
Section 8.1.4.3 has been revised to include addition details regarding penetrant testing
acceptance criteria as follows:
Liquid penetrant examined per ASME B&PV Section V, Article 6. Acceptance per
Section III, Article NF-5350 .
Page 17 of 18
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NAC INTERNATIONAL RESPONSE TO
REQUEST FOR ADDITIONAL INFORMATION
ACCEPTANCE AND MAINTENANCE TESTS EVALUATION
NAC-STC Docket No: 71-9235
Coe No: 9235
8.8 Provide justification for replacing Viton 0-rings at least once every 2 years, or revise Section 8.1.4.2 of the SAR to specify a replacement frequency of at least 1 year during
transport operations or prior to transport if they have been installed longer than 1 year.
The applicant removed reference to PTFE 0-rings in SAR Section 8.1.4.2. Due to this change, the third paragraph in the section now states: "Viton 0-rings shall be replaced at least once every two years during cask transport operations, or prior to transport if they have been installed longer than two years." The SAR Section 8.1.4.2 revision 18 from
March 2017 stated, "The Viton 0-rings shall be replaced at least annually during cask
transpo1i operations, or prior to transport if they have been installed longer than one year (i.e., for extended cask out of service periods)." Replacing elastomeric seals at an interval
not to exceed one year is consistent with Section 8.3.4.3, "Component Tests," of
NUREG-1617, "Standard Review Plan for Transportation Packages for Spent Nuclear Fuel."
This information is needed to determine compliance with 10 CFR 71.43(f) and 10 CFR 71.51(a).
NAC International Response to Acceptance and Maintenance Tests Evaluation RAI 8.8:
SAR section 8.1.4.2, third paragraph has been revised to read as follows:
"Those Vi ton 0-rings that provide the Containment Boundary seal shall be replaced annually during cask transport operations, or prior to transport of they have been installed longer than one year. Secondary Boundary (i.e., Non-Containment Boundary) Viton 0-rings shall be replaced at least once every two years during cask transport operations, or prior to transport if they have been installed longer than two years."
Page 18 of 18
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Enclosure 2 to ED20180063 Page I of2
Enclosure 2
List of Drawing Changes
NAC-STC SAR, Revision 18B
June 2018
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Enclosure 2 to ED20180063 Page 2 of2
List of Drawing Changes, NAC-STC SAR, Revision 18B
Drawing 423-209, Revision 02
1. Zone B6, Revise dimension to "47.00 +.02/-.20 TYP", was "47.00". 2. Zone F2/3, Revise dimension to "076.0", was "076.00". 3. Zone F2, Revise dimension to "0124.00 +.20/-.02", was "0124.00". 4. Zone E5, Revise dimension to "044.00 +.20/-.02", was "044.00".
Drawing 423-210, Revision 02
1. Zone F2, Revise dimension to "076.0", was "076.00". 2. Zone F2, Revise dimension to "0124.00 +.20/-.02", was "0124.00". 3. Zone E5, Revise dimension to "044.00 +.20/-.02", was "044.00".
Drawing 423-870, Rev 8
1. Zone El, removed leader with text "See Note 2" .
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Enclosure 3 to ED20180063 Page I of2
Enclosure 3
Supporting Calculations for
NAC-STC SAR, Revision 18B
June 2018
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Enclosure 3 to ED20180063 Page 2 of2
List of Calculations and Supporting Documents
I. Calculation 423-3000, Revision 6
Calculation withheld in its entirety per IO CFR 2.390 .
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Enclosure 4 to ED20180063 Page 1 of2
Enclosure 4
Proposed Changes for Certificate of Compliance Revision 19
NAC-STC SAR, Revision 18B
June 2018
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Enclosure 4 to ED20180063 Page 2 of2
CoC Sections (revised)
Page 5 of 19 5.(a)(3) Drawings
(i) The cask is constructed and assembled in accordance with the following Nuclear Assurance Corporation (now NAC International) Drawing Nos.:
Page 6 of 19
423-800, sheets 1-3, Rev. 19P & 19NP 423-802, sheets 1-7, Rev. 25 423-803, sheets 1-2, Rev. 14 423-804, sheets 1-3, Rev. 12 423-805, sheets 1-2, Rev. 8 423-806, sheets 1-2, Rev. 13 423-807, sheets 1-3, Rev. 5
5.(a)(3) Drawings
423-811, sheets 1-2, Rev. 13 423-812, Rev. 7 423-900, Rev. 8 423-209, Rev. 2 423-210, Rev. 2 423-901, Rev. 3
(ii) For the directly loaded configuration, the basket is constructed and assembled in accordance with the following Nuclear Assurance Corporation (now NAC International) Drawing Nos.:
Page 7 of 19
423-870, Rev. 8 423-871, Rev. 5 423-872, Rev. 6 423-873, Rev. 2
5.(a)(3) Drawings (Continued)
423-874, Rev. 3 423-875, sheets 1-2, Rev. 11 423-878, sheets 1-2, Rev. 5 423-880, Rev. 2P & lNP
(v) The Balsa Impact Limiters are constructed and assembled in accordance with the following NAC International Drawing Nos.:
423-257, Rev. 3 423-258, Rev. 3
423-843, Rev. 6 423-859, Rev. 1
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Enclosure 5 to ED20180063 Page I of2
Enclosure 5
List of Changes
NAC-STC SAR, Revision 18B
June 2018
• Enclosure 5 to ED20180063 Page 2 of2
List of Changes, NAC-STC SAR, Revision 18B
Chapter 1
• Page 1-v, modified List of Drawings to reflect drawing revision.
Chapter 2
• Pages 2.3.2-2 thru 2.3.2-4, updated Table Notes for Tables 2.3.2-1, 2.3.2-2 and 2.3.2-3.
Chapter 3
• No changes.
Chapter 4
• No changes.
Chapter 5
• No changes.
Chapter 6
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Chapter 7
• No changes.
Chapter 8
• Page 8.1-10, modified the third paragraph of Section 8.1.4.2. • Page 8.1-11, modified Item 3 of Section 8.1.4.3. • Page 8.1-12, modified the second paragraph of Section 8.1.5.1. • Pages 8.2-4 thru 8.2-5, modified Section 8.2.6. • Page 8.2-7, modified last row of Table 8.2-1.
Chapter 9
• No changes .
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Enclosure 6 to ED20180063 Pagel of I
Enclosure 6
SAR Changed Pages and LOEP
NAC-STC SAR, Revision 18B
June 2018
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• ANAC fdFI I NTE RNATIO NAL
Atlanta Corporate Headquarters: 3950 East Jones Bridge Road, Norcross, Georgia 30092 USA Phone 770-447-1144, Fax 770-447-1797, www.nacintl.com
June 2018
Revision l8B
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NAC-STC SAR Docket No. 71-9235
June 2018 Revision l 8B
List of Effective Pages
Chapter 1 Page 2.4-1 ..................................... Revision 18
Page 2.4.1-1 .................................. Revision 18
Page 1-i thru I-iv .......................... Revision 18 Page 2.4.2-1 .................................. Revision 18
Page 1-v ...................................... Revision l 8B Page 2.4.3-1 .................................. Revision 18
Page I-vi thru I-ix ........................ Revision 18 Pages 2.4.4-1 thru 2.4.4-6 ............. Revision 18
Pages 1-1 thru 1-12 ....................... Revision 18 Pages 2.4.4-7 thru 2.4.4-8 ........... Revision l 8B
Pages 1.1-1 thru 1.1-3 ................... Revision 18 Page 2.4.4-9 .................................. Revision 18
Page 1.1-4 ................................... Revision l 8B Pages 2.4.4-10 ............................. Revision l 8B
Pages 1.1-5 thru 1.1-46 ................. Revision 18 Page 2.4.5-1 .................................. Revision 18
Pages 1 .2-1 thru 1.2-8 ................... Revision 18 Page 2.4.6-1 .................................. Revision 18
Page 1.2-9 ................................... Revision l 8B Pages2.5.l-l thru2.5.l-38 ........... Revision 18
Pages 1.2-10 thru 1.2-42 ............... Revision 18 Pages 2.5.2-1 thru 2.5.2-29 ........... Revision 18
Pages 1.2-43 thru 1.2-44 ............. Revision l 8B Pages 2.6-1 thru 2.6-2 ................... Revision 18
Pages 1.2-45 thru 1.2-49 ............... Revision 18 Pages 2.6.1-1 thru 2.6.1-7 ............. Revision 18
Page 1.3-1 ..................................... Revision 18 Pages 2.6.2-1 thru 2.6.2-8 ............. Revision 18
Pages 1.4-1 thru 1.4-24 ................. Revision 18 Page 2.6.3-1 .................................. Revision 18
Page 2.6.4-1 .................................. Revision 18
Chapter 2 Pages 2.6.5-1 thru 2.6.5-2 ............. Revision 18
Page 2.6.6-1 .................................. Revision 18
Pages 2-i thru 2-lxviii .................... Revision 18 Page 2.6.7-1 .................................. Revision 18
Page 2-1 ........................................ Revision 18 Pages2.6.7.l-l thru 2.6.7.1-17 .... Revision 18
Pages2.1.1-l thru2.l.l-2 ............. Revision 18 Pages 2.6.7.2-1 thru 2.6.7.2-19 ..... Revision 18
Pages 2.1.1-3 thru 2.1.1-4 ........... Revision l 8B Pages 2.6. 7 .3-1 thru 2.6. 7 .3-11 ..... Revision 18
Pages 2.1.1-5 ................................. Revision 18 Pages 2.6.7.4-1 thru 2.6. 7.4-59 ..... Revision 18
Pages 2.1.2-1 thru 2.1.2-5 ............. Revision 18 Pages 2.6.7.5-1 thru 2.6.7.5-4 ....... Revision 18
Pages 2.1.3-1 thru 2.1.3-15 ........... Revision 18 Pages 2.6.7.5-5 thru 2.6.7.5-8 ..... Revision l 8B
Pages 2.2-1 thru 2.2-8 ................... Revision 18 Pages 2.6.7.5-9 thru 2.6.7.5-13 ..... Revision 18
Pages 2.3 .1-1 thru 2.3 .1-2 ............. Revision 18 Pages 2.6.7.6-1 thru 2.6.7.6-13 ..... Revision 18
Pages 2.3.2-1 ................................. Revision 18 Pages 2.6.7.7-1 thru 2.6.7.7-5 ....... Revision 18
Pages 2.3.2-2 thru 2.3.2-4 ........... Revision l 8B Page 2.6.8-1 .................................. Revision 18
Pages 2.3.2-5 ................................. Revision 18 Page 2.6.9-1 .................................. Revision 18
Pages 2.3.3-1 thru 2.3.3-2 ............. Revision 18 Page 2.6.10-1 ................................ Revision 18
Pages 2.3.4-1 thru 2.3.4-3 ............. Revision 18 Pages 2.6.10.1-1 thru
Pages 2.3.5-1 thru 2.3.5-2 ............. Revision 18 2.6.10.1-2 ................................ Revision 18
Pages 2.3 .6-1 thru 2.3 .6-5 ............. Revision 18 Pages 2.6.10.2-1 thru
Page 2.3.7-1 .................................. Revision 18 2.6.10.2-4 ................................ Revision 18
Page 2.3.8-1 .................................. Revision 18
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NAC-STC SAR Docket No. 71-9235
June2018 Revision 18B
List of Effective Pages (continued)
Pages 2.6.10.3-1 thru Pages 2.6.13.2-1 thru
2.6.10.3-7 ................................. Revision 18 2.6.13.2-7 ................................. Revision 18
Page 2.6.11-1 .............................. Revision 18B Pages 2.6.13.3-1 thru
Pages 2.6.11.1-1 thru 2.6.13.3-4 ................................. Revision 18
2.6.11.1-4 .............................. Revision 18B Pages 2.6.13.4-1 thru
Pages 2.6.11.2-1 thru 2.6.13.4-5 ................................. Revision 18
2.6.11.2-11 ............................ Revision 18B Pages 2.6.13 .5-1 thru
Page 2.6.11.3-1 .............................. Revision 18 2.6.13 .5-2 ................................. Revision 18
Pages 2.6.12-1 thru Pages 2.6.13.6-1 thru
2.6.12-5 .................................... Revision 18 2.6.13 .6-2 ................................. Revision 18
Page 2.6.12.1-1 .............................. Revision 18 Pages 2.6.13.7-1 thru
Pages 2.6.12.2-1 thru 2.6.13. 7-2 ................................. Revision 18
2.6.12.2-5 ................................. Revision 18 Page 2.6.13.8-1 .............................. Revision 18
Pages 2.6.12.3-1 thru Page 2.6.13.9-1 .............................. Revision 18
2.6.12.3-7 ................................. Revision 18 Page 2.6.13.10-1 ............................ Revision 18
Pages 2.6.12.4-1 thru Pages 2.6.13.11-1 thru
2.6.12.4-3 ................................. Rev is ion 18 2.6.13 .11-3 ............................... Revision 18
Pages 2.6.12.5-1 thru Pages 2.6.13.12-1 thru
2 .6.12.5-3 ................................. Revision 18 2.6.13 .12-2 ............................... Revision 18
Pages 2.6.12.6-1 thru Pages 2.6.14-1 thru
2.6.12.6-2 ................................. Revision 18 2. 6 .14-8 .................................... Revision 1 8
Pages 2.6.12.7-1 thru Pages 2.6.14.1-1 thru
2.6. l 2.7-22 ............................... Revision 18 2 .6 .14 .1-2 ................................. Revision 18
Pages 2.6.12.8-1 thru Pages 2.6.14.2-1 thru
2 .6 .12.8-2 ................................. Revision 18 2.6.14.2-16 ............................... Revision 18
Pages 2.6.12.9-1 thru Pages 2.6.14.3-1 thru
2.6.12.9-11 ............................... Revision 18 2.6.14.3-3 ................................. Revision 18
Page 2.6.12.10-1 ............................ Revision 18 Pages 2.6.14.4-1 thru
Page 2.6.12.11-1 ............................ Revision 18 2.6.14.4-4 ................................. Revision I 8
Page 2.6.12.12-1 ............................ Revision 18 Pages 2.6.14.5-1 thru
Pages 2.6.12.13-1 thru 2.6.14.5-3 ................................. Revision 18
2.6. l 2.13-4 ............................... Revision 18 Page 2.6.14.6-1 .............................. Revision 18
Pages 2.6.13-1 thru Pages 2.6.14.7-1 thru
2.6.13-3 .................................... Revision 18 2.6.14. 7-14 ............................... Revision 18
Pages 2.6.13.1-1 thru Pages 2.6.14.8-1 thru
2.6.13.1-2 ................................. Revision 18 2 .6.14.8-6 ................................. Revision 18
Page 2 .6.14.9-1 .............................. Revision 18
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NAC-STC SAR Docket No. 71-9235
June 2018 Revision 18B
List of Effective Pages ( continued)
Page 2.6.14.10-1 ........................... Revision 18 Pages 2.6. 16.6-1 thru
Pages 2.6.14.11-1 thru 2.6.16.6-3 ................................ Revision 18
2.6.14.11-5 .............................. Revision 18 Pages 2.6.16.7-1 thru
Pages 2.6.14.12-1 thru 2.6.16.7-12 .............................. Revision 18
2.6.14.12-5 .............................. Revision 18 Pages 2.6.16.8-1 thru
Page 2.6.15-1 ................................ Revision 18 2.6.16.8-7 ................................ Revision 18
Pages 2.6.15.1-1 thru Page 2.6.16.9-1 ............................. Revision 18
2.6.15.1-2 ................................ Revision 18 Page 2.6.16. 10-1 ........................... Revision 18
Pages 2.6.15 .2-1 thru Pages 2.6.16. 11-1 thru
2.6.15 .2-7 ................................ Revision 18 2.6.16.11-4 .............................. Revision 18
Pages 2.6.15.3-1 thru Pages 2.6.16.12- 1 thru
2.6.15.3-4 ................................ Revision 18 2.6.16.12-2 .............................. Revision 18
Pages 2.6.15.4-1 thru Pages 2.6.16.13-1 thru
2.6.15 .4-4 ................................ Revision 1 8 2.6.16.13-2 .............................. Revision 18
Page 2.6.15.5-1 ............................. Revision 18 Page 2.6.16.14-1 ........................... Revision 18
Pages 2.6.15.6-1 thru Pages 2.6.17-1 thru
2.6. 15 .6-3 ................................ Revision 18 2.6.17-13 ................................. Revision 18
Page2.6.15.7-1 ............................. Revision 18 Pages 2.6.18-1 thru
Page2.6.15.8-1 ............................. Revision 18 2.6.18-6 ................................... Revision 18
Page2.6.15.9-1 ............................. Revision 18 Pages 2.6.19-1 thru
Page 2.6.15.10-1 ........................... Revision 18 2.6.19-23 ................................. Revision 18
Pages2.6.15.11-l thru Pages 2.6.20-1 thru
2.6.15.11-3 .............................. Revision 18 2.6.20-20 ................................. Revision 18
Pages 2.6.15.12-1 thru Pages 2.6.21-1 thru
2.6.15 .12-2 .............................. Revision 18 2.6.21.-2 .................................. Revision 18
Pages 2.6.16-1 thru Pages2.7-1 thru2.7-2 ................... Revision 18
2.6.16-6 ................................... Revision 18 Page 2.7.1-1 thru 2.7.1-2 ............... Revision 18
Pages 2.6.16.1-1 thru Pages 2.7.1.1-1 thru
2.6.16.1-2 ................................ Revision 18 2.7.1.1-15 ................................ Revision 18
Pages 2.6.16.2-1 thru Pages 2.7.1.2-1 thru
2.6. 16.2-11 .............................. Revision 18 2.7.1.2-15 ................................ Revision 18
Pages 2.6.16.3-1 thru Pages 2.7.1.3-1 thru
2.6.16.3-3 ................................ Revision 18 2.7. 1 .3-9 .................................. Revision 18
Pages 2.6.16.4-1 thru Pages 2.7.1.4-1 thru
2.6.16.4-3 ................................ Revision 18 2.7.1 .4-11 ................................ Revision 18
Pages 2.6.16.5-1 thru Pages 2.7.1.5-1 thru
2.6.15.5-3 ................................ Revision 18 2.7.1.5-3 .................................. Revision 18
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NAC-STC SAR Docket No. 71-9235
June 2018 Revision I 8B
List of Effective Pages (continued)
Pages 2.7.1.6-1 thru
2.7. l .6-16 ................................. Revision 18
Page 2.7.2-1 ................................... Revision 18
Pages 2.7.2.1-1 thru
2. 7 .2.1-5 ................................... Revision 18
Pages 2.7.2.2-1 thru
2. 7 .2 .2-9 ................................... Revision 18
Pages 2.7.2.3-1 thru
2. 7 .2.3-6 ................................... Revision 18
Pages 2.7.2.4-1 thru
2.7 .2.4-7 ................................... Revision 18
Page 2.7.2.5-1 ................................ Revision 18
Page 2.7.2.6-1 ................................ Revision 18
Page 2.7.3.1-1 ................................ Revision 18
Pages 2.7.3.2-1 thru 2.7.3.2-5 ........ Revision 18
Pages 2.7.3.3-1 thru 2.7.3.3-3 ........ Revision 18
Pages 2.7.3.4-1 thru 2.7.3.4-2 ........ Revision 18
Page 2.7.3.5-1 ................................ Revision 18
Page 2.7.3.6-1 ................................ Revision 18
Page 2. 7.4-1 ................................... Revision 18
Page 2. 7 .5-1 ................................... Revision 18
Page 2. 7 .6-1 ................................... Revision 18
Pages 2.7.7-1 thru 2.7.7-4 .............. Revision 18
Pages 2.7.8-1 thru 2.7.8-4 .............. Revision 18
Pages 2.7.8. 1-1 thru 2.7.8.1-43 ...... Revision I 8
Pages 2.7.8.2-1 thru 2.7.8.2-2 ........ Revision 18
Pages 2.7.8.3-1 thru 2.7.8.3-13 ...... Revision 18
Pages 2.7.8.4-1 thru 2.7.8.4-1 O ...... Revision 18
Page 2.7.8.5-1 ................................ Revision 18
Pages 2.7.9-1 thru 2.7.9-40 ............ Revision 18
Pages 2.7.10-1 thru 2.7.10-12 ........ Revision 18
Pages2.7.11-1 thru2.7.ll-16 ........ Revision 18
Pages 2.7.12-1 thru 2.7.12-10 ........ Revision 18
Pages 2.7.13-1 thru 2.7.13-4 .......... Revision 18
Pages 2.7.13.2-1 thru
2. 7 .13 .2-2 ................................. Revision 18
Pages 2.7.13.3-1 thru
2.7.13.3-4 ................................. Revision 18
Pages 2.7.13.4-1 thru
2. 7 .13 .4-8 ................................. Revision 18
Pages 2.7.13.5-1 thru
2.7.13.5-2 ................................. Revision 18
Pages 2.7.14-1 thru
2.7.14-13 .................................. Revision 18
Pages 2.7.15-1 thru
2. 7 .15-16 .................................. Revision 18
Page 2.8-1 ...................................... Revision 18
Pages 2.9-1 thru 2.9-11 .................. Revision 18
Pages 2.10.1-1 thru 2.10.1-4 .......... Revision 18
Pages 2.10.2-1 thru 2.10.2-93 ........ Revision 18
Pages 2.10.3-1 thru 2.10.3-7 .......... Revision 18
Pages 2.10.4-1 thru
2. I 0.4-288 ................................ Revision 18
Pages 2.10.5-1 thru 2.10.5-22 ........ Revision 18
Pages 2.10.6-1 thru 2.10.6.-36 ....... Revision 18
13 drawings in Sections
2. I 0.6.6 and 2.10.6.7
Pages 2.10.6-37 thru
2. l O .6-8 8 .................................. Revision 18
Pages 2.10.7-1 thru 2.10.7-26 ........ Revision 18
Pages 2.10.8-1 thru 2.10.8-24 ........ Revision 18
Pages 2.10.9-1 thru 2.10.9-11 ........ Revision 18
Pages 2.10.10-1 thru
2.10.10-11 ................................ Revision 18
Pages 2.10.11-1 thru
2.10.11-8 .................................. Revision 18
Pages 2.7.13.1-1 thru Pages 2.10.12-1 thru
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2.7.13.1-18 ............................... Revision 18 2.10.12-31 ................................ Revision 18 •
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NAC-STC SAR Docket No. 71-9235
June2018 Revision 18B
List of Effective Pages (continued)
4 drawings in Section 2.10.12 Chapter 3
Pages 2.11.1-1 thru 2.11.1-2 ......... Revision 18 Page 3-i ....................................... Revision 18B
Pages 2.11.2-1 thru 2.11.2-2 ......... Revision 18 Pages 3-ii thru 3-iii ........................ Revision 18
Page 2.11.3-1 ................................ Revision 18 Pages 3-iv thru 3-v ...................... Revision 18B
Page 2.11.4-1 ................................ Revision 18 Page 3-vi ....................................... Revision 18
Page 2.11.5-1 ................................ Revision 18 Page 3-vii .................................... Revision 18B
Pages 2.11.6-1 thru 2.11.6-6 ......... Revision 18 Page 3-viii ..................................... Revision 18
Page 2.11.6.12-1 thru Pages 3.1-1 thru 3.1-12 ................. Revision 18
2.ll.6.12-62 ............................ Revision 18 Pages 3 .2-1 thru 3 .2-14 ................. Revision 18
Pages 2.11.6.13-1 thru Pages 3.3-1 thru 3.3-6 ................... Revision 18
2.11.6.13-35 ............................ Revision 18 Pages 3 .4-1 thru 3 .4-44 ................. Revision 18
Pages 2.11.6.14-1 thru Pages 3 .4-45 thru 3 .4-86 ............. Revision 18B
2.11.6.14-10 ............................ Revision 18 Pages 3.5-1 thru 3.5-16 ................. Revision 18
Page 2.11.6.15-1 ........................... Revision 18 Page 3 .6-1 ..................................... Revision 18
Pages 2.11. 7-1 thru 2.11. 7-8 ......... Revision 18 Pages 3.6.1-1 thru 3.6.1-4 ............. Revision 18
Pages 2.11.7.8-1 thru Pages 3 .6.2-1 thru 3 .6.2-3 ............. Revision 18
2.11. 7 .8-34 .............................. Revision 18 Pages 3.6.3-1 thru 3.6.3-3 ............. Revision 18
Pages 2.11.7.9-1 thru Pages 3 .6.4-1 thru 3 .6.4-24 ........... Revision 18
2.ll.7.9-14 .............................. Revision 18 Pages 3.6.5-1 thru 3.6.5-3 ............. Revision 18
Pages 2.11.7 .10-1 thru Page 3.7-1 ..................................... Revision 18
2.11.7.10-5 .............................. Revision 18 Pages 3.7.1-1 thru 3.7.1-3 ............. Revision 18
Page 2.11.8-1 ................................ Revision 18 Pages 3.7.2-1 thru 3.7.2-2 ............. Revision 18
Pages 2.11.9-1 thru 2.11.9-10 ....... Revision 18 Pages 3.7.3-1 thru 3.7.3-2 ............. Revision 18
Page2.12.l-l ................................ Revision 18 Pages 3.7.4-1 thru 3.7.4-9 ............. Revision 18
Pages 2.12.2-1 thru 2.12.2-2 ......... Revision 18 Page 3.7.5-1 thru 3.7.5-2 ............... Revision 18
Page 2.12.3-1 ................................ Revision 18 Page 3.8-1 ..................................... Revision 18
Page 2.12.4-1 ................................ Revision 1 8 Pages 3.8.1-1 thru 3.8.1-4 ............. Revision 18
Page 2.12.5-1 ................................ Revision 18 Pages 3.8.2-1 thru 3.8.2-3 ............. Revision 18
Page 2.12.6-1 thru 2.12.6-29 ......... Revision 18 Pages 3.8.3-1 thru 3.8.3-3 ............. Revision 18
Page 2.13.1-1 ................................ Revision 18 Pages 3.8.4-1 thru 3.8.4-17 ........... Revision 18
Pages 2.13.2-1 thru 2.13.2-2 ......... Revision 18 Pages 3.8.5-1 thru 3.8.5-2 ............. Revision 18
Page 2.13.3-1 ................................ Revision 18 Page 3.8.6-1 .................................. Revision 18
Page 2.13.4-1 ................................ Revision 18
Page 2.13 .5-1 ................................ Revision 18
Pages 2.13.6-1 thru 2.13.6-62 ....... Revision 18
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NAC-STC SAR Docket No. 71-9235
June 2018 Revision l 8B
List of Effective Pages (continued)
Chapter 4 Pages 5.7.6-1 thru 5.7.6-22 ............ Revision 18
Page 5 .8-1 ...................................... Revision 18
Page 4-i thru 4-iii ........................... Revision 18 Pages 5.8.1-1 thru 5.8.1-9 .............. Revision 18
Pages 4.1-1 thru 4.1-10 .................. Revision 18 Pages 5.8.2-1 thru 5.8.2-7 .............. Revision 18
Pages 4.2-1 thru 4.2-18 .................. Revision 18 Pages 5.8.3-1 thru 5.8.3-6 .............. Revision 18
Pages 4.3-1 thru 4.3-4 .................... Revision 18 Pages 5.8.4-1 thru 5.8.4-3 .............. Revision 18
Page 4.4-1 ...................................... Revision 18 Pages 5.8.5-1 thru 5.8.5-4 .............. Revision 18
Pages 4.5-1 thru 4.5-13 .................. Revision 18 Pages 5.8.6-1 thru 5.8.6-5 .............. Revision 18
Page 4.5-14 ................................. Revision l8B Pages 5.8.7-1 thru 5.8.7-5 .............. Revision 18
Pages 4.5-15 thru 4.5-35 ................ Revision 18 Pages 5.8.8-1 thru 5.8.8-5 .............. Revision 18
Pages 4.6-1 thru 4.6-2 .................... Revision 18 Pages 5.8.9-1 thru 5.8.9-26 ............ Revision 18
Pages 4.7-1 thru 4.7-3 .................... Revision 18
Chapter 6
Chapter 5
Pages 6-i thru 6-ix ......................... Revision 18
Page 5-i thru 5-xix ......................... Revision 18 Pages 6.1-1 thru 6.1-6 .................... Revision 18
Pages 5-1 thru 5-4 .......................... Revision 18 Pages 6.2-1 thru 6.2-11 .................. Revision 18
Pages 5 .1-1 thru 5.1-30 .................. Revision 18 Pages 6.3-1 thru 6.3-10 .................. Revision 18
Pages 5 .2-1 thru 5 .2-40 .................. Revision 18 Pages 6.4-1 thru 6.4-2 .................... Revision 18
Pages 5.3-1 thru 5.3-33 .................. Revision 18 Page 6.4.1-1 ................................... Revision 18
Pages 5 .4-1 thru 5 .4-3 .................... Revision 18 Pages 6.4.2-1 thru 6.4.2-11.. .......... Revision 18
Page 5 .4-4 ................................... Revision l 8B Pages 6.4.3-1 thru 6.4.3-29 ............ Revision 18
Pages 5.4-5 thru 5.4-41 .................. Revision 18 Pages 6.4.4-1 thru 6.4.4-30 ............ Revision 18
Pages 5.5-1 thru 5.5-61.. ................ Revision 18 Pages 6.5-1 thru 6.5-2 .................... Revision 18
Page 5 .6-1 ...................................... Revision 18 Pages 6.5.1-1 thru 6.5.1-21.. .......... Revision 18
Pages 5 .6.1-1 thru 5 .6.1-9 .............. Revision 18 Pages 6.5.2-1 thru 6.5.2-20 ............ Revision 18
Pages 5.6.2-1 thru 5.6.2-20 ............ Revision 18 Pages 6.6-1 thru 6.6-2 .................... Revision 18
Pages 5.6.3-1 thru 5.6.3-13 ............ Revision 18 Pages 6.7-1 thru 6.7-333 ................ Revision 18
Pages 5.6.4-1 thru 5.6.4-34 ............ Revision 18 Page 6.8-1 ...................................... Revision 18
Page 5.6.5-1 ................................... Revision 18 Pages 6.8.1-1 thru 6.8.1-6 .............. Revision 18
Pages 5.6.6-1 thru 5.6.6-57 ............ Revision 18 Pages 6.8.2-1 thru 6.8.2-2 .............. Revision 18
Page 5.7-1 ...................................... Revision 18 Pages 6.8.3-1 thru 6.8.3-20 ............ Revision 18
Pages 5.7.1-1 thru 5.7.1-5 .............. Revision 18 Pages 6.8.4-1 thru 6.8.4-34 ............ Revision 18
Pages 5.7.2-1 thru 5.7.2-5 .............. Revision 18 Pages 6.8.5-1 thru 6.8.5-34 ............ Revision 18
Pages 5.7.3-1 thru 5.7.3-10 ............ Revision 18 Page 6.8.6-1 ................................... Revision 18
Pages 5.7.4-1 thru 5.7.4-14 ............ Revision 18 Pages 6.8.7-1 thru 6.8.7-27 ............ Revision 18
Page 5. 7 .5-1 ................................... Revision 18 Page 6.9-1 ...................................... Revision 18
6 of7
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NAC-STC SAR Docket No. 71-9235
List of Effective Pages ( continued)
Page 6.9 .1-1 .................................. Revision 18
Page 6.9.2-1 .................................. Revision 18
Chapter 7
Pages 7-i thru 7-ii .......................... Revision 18
Page 7-1 thru 7-3 ........................... Revision 18
Pages 7 .1-1 thru 7.1-18 ................. Revision 18
Pages 7 .2-1 thru 7 .2-5 ................... Revision 18
Pages 7.3-1 thru 7.3-10 ................. Revision 18
Pages 7.4-1 thru 7.4-11 ................. Revision 18
Page 7 .5-1 ..................................... Revision 18
Pages 7 .6-1 thru 7 .6-6 ................... Revision 18
Chapter 8
Page 8-i thru 8-ii ......................... Revision 18B
Page 8-1 ........................................ Revision 18
Pages 8.1-1 thru 8.1-5 ................... Revision 18
Pages 8.1-6 thru 8.1-7 ................. Revision 18B
Pages 8.1-8 thru 8.1-9 ................... Revision 18
Pages 8.1-10 thru 8.1-25 ............. Revision 18B
Pages 8.1-26 thru 8.1-38 ............... Revision 18
Pages 8.2-1 thru 8.2-3 ................... Revision 18
Pages 8.2-4 thru 8.2-7 ................. Revision 18B
Page 8.3-1 ..................................... Revision 18
Pages 8.4-1 thru 8.4-3 ................. Revision 18B
Page 8.4-4 ..................................... Revision 18
Pages 8.4-5 thru 8.4-14 ............... Revision 18B
Chapter 9
Page 9-i ......................................... Revision 18
Pages 9-1 thru 9-13 ....................... Revision 18
7 of7
June 2018 Revision 18B
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NAC-STC SAR
Docket No. 71-9235, Revisions 17C and 18A
June2018
Revision 18B
List of Drawings
Revision Drawing Number No. Title
423-800, sheets 1-3 Rev 19pO) Cask Assembly - NAC-STC Cask
423-800, sheets 1-3 Rev 19NP0l Cask Assembly- NAC-STC Cask
423-802, sheets 1-7 Rev 25 Cask Body - NAC-STC Cask
423-803, sheets 1-2 Rev 14 Lid Assembly - Inner, NAC-STC Cask
423-804, sheets 1-3 Rev 12 Details - Inner Lid, NAC-STC Cask
423-805, sheets 1-2 Rev 8 Lid Assembly - Outer, NAC-STC Cask
423-806, sheets 1-2 Rev 13 Port Coverplate Assy - Inner Lid, NAC-STC Cask
423-807, sheets 1-3 Rev 5 Assembly, Port Cover, NAC-STC Cask
423-209 Rev2 Impact Limiter Assy - Upper, NAC-STC Cask
423-210 Rev2 Impact Limiter Assy - Lower, NAC-STC Cask
423-257 Rev 3 Balsa Impact Limiter, Upper, NAC-STC Cask
423-258 Rev 3 Balsa Impact Limiter, Lower, NAC-STC Cask
423-811, sheets 1-2 Rev 13 Details - NAC-STC Cask
423-812 Rev 7 Nameplates -NAC-STC Cask
423-843 Rev 6 Transport Assembly, Balsa Impact Limiters, NAC-STC
423-859 Rev 1 Attachment Hardware, Balsa Limiters, NAC-STC
423-870 Rev 8 Fuel Basket Assembly, PWR, 26 Element, NAC-STC Cask
423-871 Rev 5 Bottom Weldment, Fuel Basket, PWR, 26 Element, NAC-STC Cask
423-872 Rev 6 Top Weldment, Fuel Basket, PWR, 26 Element, NAC-STC Cask
423-873 Rev 2 Support Disk and Misc. Basket Details, PWR, 26 Element, NAC-STC Cask
423-874 Rev 3 Heat Transfer Disk, Fuel Basket, PWR, 26 Element, NAC-STC Cask
423-875, sheets 1-2 Rev 11 Tube, NAC-STC Cask
423-878, sheets 1-2 Rev 5 Alternate Tube Assembly, NAC-STC Cask
423-880 Rev 2p(I) Shielded Thermal Shunt Assembly, NAC-STC Cask
423-880 Rev lNP(I) Shielded Thermal Shunt Assembly, NAC-STC Cask
423-900 Rev 8 Package Assembly Transportation, NAC-STC Cask
423-901, sheets 1-2 Rev 3 Transportation Package Concept, NAC-STC Cask
455-800, sheets 1-2 Rev 2 Assembly, Transport Cask, MPC-Yankee
(I) Proprietary and Non-proprietary drawing versions are only included in their respective SAR versions .
1-v
NAC-STC SAR
Docket No. 71-9235
Drawing Number 455-801, sheets 1-2 455-820, sheets 1-2
455-870
455-871, sheets 1-2
455-871, sheets 1-3
455-872, sheets 1-2
455-872, sheets 1-2
455-873
455-881, sheets 1-3
455-887, sheets 1-3
455-888, sheets 1-2
455-891, sheets 1-2
455-891, sheets 1-3
455-892, sheets 1-2
455-892, sheets 1-3
455-893
455-894
455-895, sheets 1-2
455-895, sheets 1-2
455-919
414-801, sheets 1-2
414-820
414-870
414-871, sheets 1-2
414-872, sheets 1-3
414-873
414-874
414-875
414-881, sheets 1-2
414-882, sheets 1-2
March 2017
Revision 18
List of Drawings (continued)
Revision No. Title
Rev4 Assembly, Transport Cask, NAC-MPC Rev 3 Spacers, Transpo11 Cask, MPC-Yankee
Rev 5 Canister Shell, MPC-Yankee
Rev 8 Details, Canister, MPC-Yankee
Rev 7P2 Details, Canister, MPC-Y ankee
Rev 12 Assembly, Transportable Storage Canister (TSC), MPC-Yankee
Rev 1 lPl Assembly, Transportable Storage Canister (TSC), MPC-Yankee
Rev4 Assembly, Drain Tube, Canister, MPC-Yankee
Rev 8 PWR Fuel Tube, MPC-Yankee
Rev4 Basket Assembly, 24 GTCC Container, MPC-Yankee
Rev 8 Assembly, Transportable Storage Canister (TSC), 24 GTCC Container, MPC-Yankee
Rev 1 Bottom Weldment, Fuel Basket, MPC-Yankee
Rev 2PO Bottom Weldment, Fuel Basket, MPC-Yankee
Rev 3 Top Weldment, Fuel Basket, MPC-Yankee
Rev 3PO Top Weldment, Fuel Basket, MPC-Yankee
Rev 3 Support Disk and Misc. Basket Details, MPC-Yankee
Rev 2 Heat Transfer Disk, Fuel Basket, MPC-Yankee
Rev 5 Fuel Basket Assembly, MPC-Yankee
Rev 5PO Fuel Basket Assembly, MPC-Yankee
Rev2 Retainer, United Nuclear Test Assy, MPC-Yankee
Rev 2 Cask Assembly, NAC-STC, CY-MPC
Rev 0 Canister Spacer CY-MPC
Rev 3 Canister Shell, CY-MPC
Rev 6 Details, Canister CY-MPC
Rev 6 Assembly, Transportable Storage Canister (TSC), CY-MPC
Rev 2 Drain Tube Assembly, CY-MPC
Rev 0 Shim, Canister, CY-MPC
Rev 0 Spacer Shim, Canister, CY-MPC
Rev4 Fuel Tube, Transportable Storage Canister (TSC), CY-MPC
Rev 4 Oversize Fuel Tube, Transportable Storage Canister (TSC), CY-MPC
1-vi
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NAC-STC SAR
Docket No. 71-9235
(4) Vent port coverplate interseal test hole threaded plug with metallic 0-ring;
(5) Drain port coverplate and the coverplate outer metallic 0-ring; and,
(6) Drain port coverplate interseal test hole threaded plug with metallic 0-ring.
March 2017
Revision 18
Metallic 0-rings are required for the storage configuration and are qualified for transport prior to
shipment in accordance with the operating procedure.
The NAC-STC is designed to meet IO CFR 71 and IAEA Safety Series No. SSR-6 licensing
requirements for spent fuel transport packages. The transport licensing requirements include
providing safe containment during the handling and transpo1t of spent nuclear fuel. Certain design
features of the NAC-STC that have been included for the sole purpose of satisfying storage
licensing requirements also provide added safety for transport conditions. The design features of
the NAC-STC include: inner and outer lids, redundant seals at each containment boundary
penetration, cavity penetrations located in the inner lid, and a puncture-resistant outer shell and
outer lid.
This Safety Analysis Report is written for transport cask licensing only. Design features related
to storage cask licensing are included for clarity and for ease of review.
The NAC-STC closure design provides dual lids for transport and storage operations, as well as
protection of the vent and drain ports that are located in the inner lid. This design permits
performance of a periodic verification leak test on the containment seals prior to transpo1t
following extended storage. Both the inner and outer lids are installed during transport and storage.
The inner lid and its 0-rings are the major removable components in the primary containment
boundary. Two concentric 0-rings are used to seal the inner lid to the cask cavity flange. An
0-ring test port connects to the annulus between the two 0-rings to permit leak testing.
The vent and drain port coverplates, which protect the vent and drain ports located in the inner lid,
are also part of the primary containment boundary of the cask. Each coverplate is sealed by two
concentric 0-rings.
As described in Section 4. I, the inner 0-rings of the inner lid and two coverplates are the
containment boundary for contents ( either directly loaded fuel or a loaded transportable storage
canister) that is loaded for transport without interim site storage. The outer 0-rings of these
I. 1-3
NAC-STC SAR
Docket No. 71-9235, Revision 17C
June 2018
Revision 18B
components are the containment boundary for directly loaded fuel that is to be transported after an
extended period of storage.
The inner lid and coverplate 0-rings may be either metallic or non-metallic as shown in the License
Drawings. However, metallic 0-rings must be used when the NAC-STC is directly loaded for
long-term storage or for the transport of canistered contents. The metallic 0-rings provide long
term sealing capability in an elevated temperature and radiation environment.
The outer lid provides a sealed secondary closure for transport and storage operations using a
single 0-ring. The 0-ring may be either metallic or non-metallic. The outer lid protects the inner
lid and the vent and drain ports from external puncture events.
There are two penetrations in the top forging: an interlid port, which serves primarily as a drain
for the inter! id region, and a pressure port, which may house a transducer that monitors the pressure
in the interlid region during storage. During transpo1t, the pressure port is closed by a threaded
plug. The pressure port plug is covered by the transport p01t coverplate. The interlid and pressure
port penetrations in the top forging are protected by SA-705, Type 630, 17-4 precipitation
hardened (PH) stainless steel po1t covers with two Viton 0-rings.
The body of the NAC-STC is a smooth right-circular cylinder of multiwall construction, consisting
of stainless steel inner and outer shells separated by lead gamma radiation shielding, which is
poured in place. The center section of the inner shell is fabricated from Type 304 stainless steel.
At each end of the inner shell center section, inner shell rings fabricated from Type XM-19
stainless steel provide the transition to the bottom inner forging and the top forging. The outer shell
is also fabricated from Type 304 stainless steel. The inner and outer shells are welded to the Type
304 stainless steel top forging, which is a ring that is machined to mate with the inner and outer
lids. The inner and outer shells are also welded to the Type 304 stainless steel bottom inner and
outer forgings, respectively. The cask bottom consists of the two forgings and a plate with neutron
shield material sandwiched between the bottom inner forging and the bottom plate. Neutron shield
material is also placed in an annulus that surrounds the cask outer shell along the length of the cask
cavity. The neutron shielding material is a solid synthetic polymer (NS-4-FR). The neutron shield
annulus is enclosed by a Type 304 stainless steel shell and by end plates that are welded to the
outer shell. Two pressure relief valves are provided in the bottom of the neutron shield annulus to
relieve pressure in the neutron shield annulus due to a severe thermal accident condition (fire).
Neutron shielding is also provided on the top of the cask by a layer ofNS-4-FR enclosed in the
inner lid.
1.1-4
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NAC-STC SAR
Docket No. 71-9235, Revision l 7C
June 2018
Revision 18B
The pressure port, located in the top forging, houses the transducer that monitors the pressure in the
interlid region during storage. The pressure transducer is removed during transpo1t. The interlid
po11 penetrates the top forging into the region between the inner and the outer lids and serves as a
drain for the interlid region and as a port to pressurize the interlid region for seal testing purposes.
The inter! id port is closed by a quick disconnect. The basic geometry of the inter! id and pressure
ports and port covers is identical. Each po11 has a 4.5-inch diameter opening that is a minimum of
1.1 inches deep. Concentric with the port opening is a 2.93-inch diameter bore. This bore acts as a
lead-in to the 2.875-inch diameter bore that serves as the sealing surface for the two Viton 0-rings
in the po11 cover.
Both of the port covers are fabricated from SA-705, Type 630, HJ 150, 17-4 PH stainless steel.
The port covers resemble a cup-shape and have the geometrical appearance of a thick round end
plate with a cylindrical body. The end plate of the port cover is 4.5 inches in diameter and 1.0
inch thick. The three 3/8 - 16 UNC port cover bolts, which are fabricated from SA-193, Grade
86, Type 410 stainless steel, are countersunk flush with the top of the port cover. There are two
Viton 0-rings on the cylindrical body of the port covers with a seal test port between the 0-rings.
A retainer is bolted to the open end of the cylindrical body of the port cover to retain the 0-rings
and the spacer between them after assembly. The port cover design permits the thick end plate to
absorb an impact, while any deflection of the end plate results in the 0-rings sliding in the bore of
the port with the seal maintained.
The basic geometry of the vent port and coverplate, and the drain port and coverplate, are identical
to each other. Each port has a 6.53-inch diameter opening in the inner lid that is 1.8 inches deep.
Concentric with the port opening is a 3.25-inch diameter bore that houses the 1.0-inch diameter
quick disconnect. As shown in Drawing 423-806, the 1.0-inch thick vent and drain port
coverplates are fabricated from SA-240, Type 304 stainless steel. When installed, the port
coverplates are recessed 0.8 inch below the top surface of the inner lid. The vent and drain poti
coverplates are sealed to the inner lid by the metallic or nonmetallic 0-rings on the bottom face of
each port coverplate. The four 1 /2 - 13 UNC port coverplate bolts are fabricated from SA-193,
Grade 86, Type 410 stainless steel. The bolt holes are countersunk so that the bolt heads are flush
with the top of the po11 coverplate. Metallic 0-rings are used for storage and for transport
following storage, and for transpo1i of canistered spent fuel and HL W without interim storage.
Either metallic or non-metallic 0-rings may be used for transport without interim storage after
loading. The outer metallic 0-ring provides the primary containment seal for transport after
storage, while the inner 0-ring (either metallic or non-metallic) provides the primary containment
seal for transport without interim storage after loading .
1.2-9
NAC-STC SAR
Docket No. 71-9235
1.2.1.2.5 Lifting Trunnions and Rotation Trunnion Recesses
March 2017
Revision 18
The NAC-STC has four lifting trunnions that are fabricated from SA-705, Type 630, Hl 150,
17-4 PH stainless steel and are welded into 2.0-inch deep recesses in the top forging at 90-degree
intervals around the cask circumference. Only two diametrically opposite lifting trunnions are
required to lift the NAC-STC. The lifting trunnions are 5.5 inches in diameter and have a
load-bearing width of 2.5 inches. The trunnions are machined to create a 0.38-inch thick end
flange, which acts as a safety stop to ensure proper engagement and to prevent inadvertent
disengagement of the lifting yoke.
There are two rotation trunnion recesses located near the bottom end of the NAC-STC. The
rotation trunnion recesses are located approximately 18 inches above the bottom of the cask in line
with two of the lifting trunnion, but 3.0 inches offset from the cask centerline to ensure that rotation
of the cask occurs in the proper direction. Each recess is fabricated from SA-705, Type 630, 17-4
PH stainless steel and is groove-welded to the bottom outer forging. The recess is 6.0 inches
square and 4.13 inches deep, with a full radius at the top of the recess that engages with the rotation
support. The neutron shield shell is cut out to accommodate the rotation trunnion recesses.
1.2.1.2.6 Transport Impact Limiters
The NAC-STC is equipped with removable, cup-shaped impact limiters that are bolted over each
end of the cask to ensure that the design impact loads for the cask are not exceeded for any of the
defined normal operation or accident drop conditions. The NAC-STC transport impact limiters
are provided in two configurations. The standard configuration is constructed of a combination of
redwood and balsa wood and is referred to as the redwood impact limiter. The other configuration
is constructed using only balsa wood and is referred to as the balsa impact limiter. Both
configurations are completely enclosed in a stainless steel shell. The upper impact limiter has
cutouts in its inside diameter for clearance with the lifting trunnions. The impact limiters absorb
the energy of a cask drop by crushing the redwood and/or balsa wood. The force required to crush
the impact limiter is determined by the amount and location of the wood and its grain direction.
The upper and lower impact limiters are bolted over each end of the cask body by 16 equally
spaced attachment rods and nuts. The lightweight impact limiters have a lower weight and
improved crush characteristics compared to the standard impact limiters, and accommodate a
higher cask content weight and higher cask total weight. As shown in Section
1.2-10
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NAC-STC SAR
Docket No. 71-9235, Revision 17C
Table 1.2-1 Design Characteristics of the NAC-STC (continued)
Design Characteristics Dimension 1
Seals (0-rings) for Spent Fuel Storage Configuration Prior to Transport, and Immediate Transport Configuration for Canister Spent Fuel, GTCC Waste, and HL W Overpacks
- Inner Lid
- Inner 0.25 dia. X n.251 dia. Metal Seal
- Outer 0.25 dia. x 73.497 dia. Metal Seal
- Port Coverplates
- Inner 0.125 dia. x 3.875 dia. Metal Seal
- Outer 0.125 dia. x 4.500 dia. Metal Seal
- Outer Lid 0.250 dia. x 82.060 dia. Metal Seal
- Port Covers
- Primary 0.103 dia. x 2.675 dia. Viton
- Secondary 0.103 dia. x 2.675 dia. Viton
1.2-43
June 2018
Revision 18B
Material
NAC-STC SAR
Docket No. 71-9235, Revision 17C
June2018
Revision 18B
Table 1.2-1 Design Characteristics of the NAC-STC (continued)
Design Characteristics Dimension 1 Material
Seals (0-rings) for Immediate Spent Fuel Transpo11 Configuration
- Inner Lid
- Inner 0.25 dia. x 72.251 dia. Viton or Metal Seal
- Outer 0.25 dia. x 73.497 dia. Viton
- Port Coverplates
- Inner 0.125 dia. x 3.875 dia. Viton or Metal Seal
- Outer 0.125 dia. x 4.500 dia. Viton
- Outer Lid 0.250 dia. x 82.060 dia. Viton
- Port Covers
- Primary 0.103 dia. x 2.675 dia. Viton
- Secondary 0.103 dia. x 2.675 dia. Viton I. Dimensions in Inches unless otherwise noted.
1.2-44
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Drawing Nos. Drawing 423-209, Revision 02; Drawing 423-210, Revision 02; and Drawing 423-870, Rev 8 have been withheld as Sensitive Unclassified Non-Safeguards Information pursuant to 10 CFR 2.390.
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NAC-STC SAR
Docket No. 71-9235, Revision I 7C
June 2018
Revision I SB
bottom connects the inner and outer shells, providing for the bottom end closure, as well as both
gamma and neutron radiation shielding in the axial direction.
The inner lid, bolts, and 0-rings are the primary closure components of the NAC-STC for
transport conditions. The outer lid and 0-ring provide a secondary closure boundary.
The vent po1t and the drain port are located in the inner lid and are each protected by a port
coverplate. The primary containment boundary at the vent port and at the drain po1t is the port
coverplate and its 0-rings. The 0-ring is located in the bottom surface of the port coverplate. A
second 0-ring is also located in the bottom surface of the port coverplate, inside of, and
concentric with, the first 0-ring.
The forty-two 1 I /2 - 8 UN inner lid bolts are preloaded by an installation torque to restrain
rotation of the edge of the inner lid and to maintain a containment seal for the critical load
condition. This condition is a uniformly distributed pressure resulting from the impact of the
basket and cavity contents on the inner surface during a top end or top corner impact. The
critical design load condition for the inner lid bolts, as listed in Table 2.7.1.6-2, Section 2.7.1.6,
is a 54.7 g top corner impact (IO CFR 71 Hypothetical Accident Condition). The critical design
load condition for the inner lid is the top end impact, Section 2.7.1.6.
The outer lid is bolted to the top forging by the thirty-six 1 - 8 UNC outer lid bolts, which are
installed to a specified torque. The torque provides a total bolt preload that exceeds the
maximum applied bolt load for the critical load condition, preventing any lid and 0-ring
movement that might result in a loss of secondary seal integrity. The critical design load
condition for the outer lid bolts, as listed in Table 2.7.1.6-4, Section 2.7.1.6, is a 51.3g side
impact (10 CFR 71 Hypothetical Accident Condition). The critical design load condition for the
outer lid is the pin puncture accident condition. The NAC-STC outer lid bolts are loaded by the
interlid region pressure, the 0-ring compression force, and by either the impact limiter crush
force during a top end or top corner impact, or by a concentrated center load during a pin
puncture impact. The outer lid seal is provided by an 0-ring, which is tested by pressurizing the
interlid region.
In addition to the main closure, the secondary closure boundary of the NAC-STC also includes
the two po1ts located in the top forging-the interlid port and the pressure port. Each of these
ports is protected and sealed by a recessed, bolted port cover with two Viton 0-rings. The port
covers are installed with new 0-rings just prior to transport (a slightly different port cover is
2.1.1-3
NAC-STC SAR
Docket No. 71-9235, Revision 17C
June 2018
Revision l 8B
installed during storage operation). The seal at each port cover is verified by pressure-testing the
annulus between the two Viton 0-rings.
The neutron shielding material, NS-4-FR, is a solid synthetic polymer that absorbs the neutron
radiation emitted by the cask contents. In addition to the radial neutron shielding along the cask
length, neutron shielding is provided in the axial direction at each end of the cask by circular
layers ofNS-4-FR enclosed in the inner lid and in the cask bottom.
Four external trunnions are welded to the top forging of the NAC-STC at 90-degree intervals
around the circumference of the cask. These trunnions are provided for lifting and handling the
cask. Either a redundant (four trunnions) or a nonredundant (two trunnions) lifting system may
be used. However, each pair of opposing trunnions are conservatively designed to satisfy the
heavy lifting requirements of NUREG-0612 for a nonredundant lift, as well as the requirements
of 10 CFR 7 l .45(a) and paragraph 607 of IAEA Safety Standards Series No. SSR-6. Two
rotation trunnion recesses are welded to the bottom outer forging near the bottom of the cask.
The neutron shield is cut out to accommodate the placement of the rotation trunnion recesses,
which are used to attach the bottom of the cask to the transport vehicle and to rotate the cask
from the vertical lifting position to the horizontal position and vice-versa.
As discussed above, two transport impact limiter configurations are used with the NAC-STC
cask to limit the impact loads that may act on the cask. The impact limiters absorb the energy of
a cask drop impact through the crushing of the wood in the limiters. A balsa impact limiter
design must be used when the NAC-STC is transporting spent fuel or GTCC waste in the
CY-MPC canister configuration. When transpo1ting directly loaded fuel or the Yankee-MPC
canister, either the redwood or balsa impact limiter configuration may be used.
The NAC-STC fuel basket is constructed of stainless steel and has a capacity of 26 PWR fuel
assemblies. The fuel basket has a cylindrical shape with a series of support disks that provide
lateral support for the square, stainless steel fuel tubes, which encase neutron absorber sheets or
plates on each of the four sides. The support disks are separated and supported at 4.87-inch
intervals by a threaded rod and spacer nuts at six locations. Aluminum heat transfer disks are
located in the central region of the fuel basket and are supported by the six threaded rods and
spacer nuts. The stainless steel support disks have adequate strength at the basket temperatures
that occur during the transport and/or storage of 26 design-basis PWR fuel assemblies.
For the Yankee Class fuel and GTCC waste, the Yankee-MPC transportable storage canister
(canister) serves as the enclosure of the spent fuel assemblies, damaged fuel cans and GTCC
2.1.1-4
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NAC-STC SAR
Docket No. 71-9235
2.3.2 Austenitic Stainless Steels
March 2017
Revision 18
The primary structural components of the NAC-STC body, excluding: (1) the inner shell rings
(transition sections of the inner shell); (2) the outer lid; (3) the lifting trunnions; and (4) the
rotation trunnion recesses, are fabricated from Type 304 stainless steel. In addition to the cask
body components fabricated from Type 304 stainless steel, the fuel tubes and fuel basket top and
bottom weldment plates are fabricated from the same material. This material is selected because
it is strong, ductile, and highly resistant to corrosion and brittle fracture. Type XM-19 stainless
steel is selected for the inner shell rings at the ends of the inner shell because the high strength of
Type XM-19 stainless steel provides additional resistance to shear buckling in those sections of
the inner shell.
The mechanical properties of SA-240 (plate), Type 304 stainless steel are tabulated in Table
2.3.2-1. The mechanical properties of SA-336 (forging), Type 304 stainless steel are tabulated in
Table 2.3.2-2. The mechanical properties of SA-240 (plate), Type XM-19 stainless steel are
tabulated in Table 2.3.2-3 .
The primary structural components of the Yankee-MPC and CY-MPC canisters and baskets,
excluding the support disks and heat transfer disks, are fabricated from Type 304 and Type 304L
stainless steels. This material is selected because it is strong, ductile, and highly resistant to
corrosion and brittle fractures. The associated mechanical properties of the Type 304 and 304L
stainless steels are tabulated in Tables 2.3.2-1, 2.3.2-2 and 2.3.2-4 .
2.3.2-1
NAC-STC SAR
Docket No. 71-9235
Table 2.3 .2-1 Mechanical Properties of SA 240, Type 304 Stainless Steel
Temperature (°F)
Property (units) -40 -20 +70 +200 +300
Ultimate Strength 1 75.0 75.0 75.0 71.0 66.0 (ksi)
Yield Strength2 30.0 30.0 30.0 25.0 22.5 (ksi)
Design Stress 20.0 20.0 20.0 20.0 20.0 Jntensity3 (ksi)
Modulus of 28.7E+3 28.7E+3 28.3E+3 27.6E+3 27.0E+3 Elasticity4 (ksi)
Alternating Stress5 718.0 718.0 708.0 690.5 675.5 (iv, IO cycles (ksi)
Alternating Stress5 28.7 28.7 28.3 27.6 27.0 (a), 106 cycles (ksi)
Coefficient of 8.13E-6 8.19E-6 8.46E-6 8.79E-6 9.00E-6 Thermal Expansion6
(in/inl°F)
Poisson's Ratio7 0.31
Density (lbm/in3 ) 497 lbm/ft3 (0.288 lbm/in3)
1 "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table U. 2 "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table Y-1. 3 "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table 2A. 4 "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table TM-I. 5 "ASME Boiler and Pressure Vessel Code," Section Ill, Appendix I, Table 1-9.1. 6 "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table TE-I. 7 "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table NF-I.
2.3.2-2
+400
64.4
20.7
18.7
26.5E+3
663.0
26.5
9.19E-6
June 2018
Revision 18B
+500 +750
63.5 63.1
19.4 17.3
17.5 15.6
25.8E+3 24.4E+3
645.5 610.4
25.8 24.4
9.37E-6 9.76E-6
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NAC-STC SAR
Docket No. 71-9235, Revision 17B
Table 2.3.2-2 Mechanical Properties of SA 336, Type 304 Stainless Steel
Temperature (°F)
Property (units) -40 -20 +70 +200 +300
Ultimate Strength 1 70.0 70.0 70.0 66.2 61.5 (ksi)
Yield Strength2 30.0 30.0 30.0 25.0 22.5 (ksi)
Design Stress 20.0 20.0 20.0 20.0 20.0 Intensity3 (ksi)
Modulus of 28.7E+3 28.7E+3 28.3E+3 27.6E+3 27.0E+3 Elasticity4 (ksi)
Alternating Stress5 718.0 718.0 708.0 690.5 675.5 @ IO cycles (ksi)
Alternating Stress5 28.7 28.7 28.3 27.6 27.0 @ I 06 cycles (ksi)
Coefficient of 8.13E-6 8.19E-6 8.46E-6 8.79E-6 9.00E-6 Thermal Expansion6
(in/in/°F)
Poisson's Ratio7 0.31
Density8 (lbm/in3) 497 lbm/ft3 (0.288 lbm/in3)
1 "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table U. 2 "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table Y-1. 3 "ASME Boiler and Pressure Vessel Code," Section II, Pmi D, Table 2A. 4 "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table TM-I.
"ASME Boiler and Pressure Vessel Code," Section III, Appendix I, Table 1-9.1. 6 "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table TE-I. 7 "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table NF-!. 8 "Nuclear Materials Handbook," Volume I, Design Data, Property Code 3304 .
2.3.2-3
+400
60.0
20.7
18.7
26.5E+3
663.0
26.5
9.19E-6
June2018
Revision 18B
+500 +750
59.3 58.9
19.4 17.3
17.5 15.6
25.8E+3 24.4E+3
645.5 610.4
25.8 24.4
9.37E-6 9.76E-6
NAC-STC SAR
Docket No. 71-9235
Table 2.3.2-3 Mechanical Properties of Type XM-19 Stainless Steel
Temperature (°F)
Property (units)9 -40 -20 +70 +200 +300 +400
Ultimate Strength 1 100.0 100.0 100.0 99.5 94.3 90.7 (ksi)
Yield Strength2 55.0 55.0 55.0 47.0 43.4 40.8 (ksi)
Design Stress 33.3 33.3 33.3 33.2 31.4 30.2 lntensity3 (ksi)
Modulus of 28.3E+3 28.3E+3 28.3E+3 27.0E+3 27.0E+3 26.5E+3 Elasticity4 (ksi)
Alternating Stress5 708.0 708.0 708.0 690.5 675.5 663.0 (]iJ, 10 cycles (ksi)
Alternating Stress5 28.3 28.3 28.3 27.6 27.0 26.5 (]iJ, 106 cycles (ksi)
Coefficient of 8.13E-6 8.l 9E-6 8.46E-6 8.79E-6 9.00E-6 9.19E-6 Thermal Expansion6
(in/in/°F)
Poisson's Ratio7 0.31
Density8 (lbm/in3) 497 lbm/ft3 (0.288 lbm/in3)
"ASME Boiler and Pressure Vessel Code," Section II, Part D, Table U. 2 "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table Y-1. 3 "ASl\!lE Boiler and Pressure Vessel Code," Section II, Part D, Table 2A. 4 "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table TM-I. 5 "ASME Boiler and Pressure Vessel Code," Section Ill, Appendix I, Table 1-9.1. 6 "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table TE-I. 7 "ASME Boiler and Pressure Vessel Code," Section II, Part D, Table NF-1. 8 "Nuclear Materials Handbook," Volume I, Design Data, Property Code 3304.
June 2018
Revision l 8B
+500 +750
89.1 85.7
38.8 35.8
29.7 28.5
25.8E+3 24.4E+3
645.5 610.4
25.8 24.4
9.37E-6 9.76E-6
9 SA-182, FXM-19 stainless steel may be substituted for SA-240 XM-19 stainless steel provided that the SA-182 material yield and ultimate strengths are equal to or greater than those of the SA-240 material. The SA-182 forging material and the SA-240 plate material are both XM-19 austenitic stainless steels. Austenitic stainless steels do not experience a ductile-to-brittle transition for the range of temperatures considered in this Safety Analysis Report. Therefore, fracture toughness is not a concern.
2.3.2-4
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NAC-STC SAR
Docket No. 71-9235, Revision 17C
2.4.4.2.3 Shielding Materials
June 2018
Revision 18B
The primary shielding materials used in the NAC-STC-lead and NS-4-FR-are completely
enclosed and sealed in stainless steel. As previously described, there are no potential reactions of
these materials with the stainless steel or with the copper fins.
Therefore, there are no potential reactions associated with the cask shielding materials.
2.4.4.2.4 Criticality Control Material
The criticality control material is a sheet consisting of boron carbide mixed m an aluminum
alloy. This material is effectively a sheet of aluminum that is in contact with the aluminum alloy
fuel tubes and is exposed to the cask cavity environment. This material is protected by an oxide
layer that formed shortly after fabrication. The existing oxide layer effectively precludes fu11her
oxidation of the aluminum. Consequently, there are no potential reactions associated with the
aluminum-based criticality control material.
2.4.4.2.5 Energy Absorbing Material
The NAC-STC utilizes redwood and balsa wood for energy absorption in the impact limiters.
The wood is completely enclosed (sealed) in stainless steel and there are no potential reactions
between the wood and the stainless steel shells. The wood may be coated with a preservative
prior to installation in the impact limiter shell and blocks of wood may be glued together with an
epoxy adhesive. These are standard applications of preservatives and adhesives, so no post
application reactions will occur.
There are no potential reactions associated with the energy absorbing material.
2.4.4.2.6 Cellular Foam and Insulation
The NAC-STC utilizes layers of expansion foam and strips of insulation in the solid neutron
shield regions. The expansion foam permits thermal expansion of the solid neutron shield
material during normal operation, and the insulation protects the expansion foam during final
closure welding of the neutron shield shell to the end plate. The foam and the insulation are
nonflammable, nontoxic and noncorrosive silicone products that are used in the casks in a
standard design application .
2.4.4-7
NAC-STC SAR
Docket No. 71-9235, Revision I 7C
June2018
Revision 18B
There are no potential reactions associated with the silicone expansion foam or insulation.
2.4.4.2.7 Lubricant and Grease
The dry film lubricants used with the NAC-STC meet the performance and general
compositional requirements of the nuclear power industry. One example is NEVER-SEEZ®
lubricant, which can be used on rotating bearing surfaces. Another example is Neolube®, which
can be used on threaded/mechanical connection surfaces. In addition, Dow Corning High
Vacuum Grease is an example of what can be used as an adherent/lubricant to lubricate and
retain the 0-ring seals in their grooves. None of these example lubricants contain elements or
compounds prohibited by the NRC. NEVER-SEEZ® is a superior, high temperature, anti-seize
and extreme pressure lubricant that contains flake particles of pure nickel, graphite and other
additives in a special grease carrier. Neolube® is 99% pure furnace graphite particles in
isopropanol. It has excellent radiation resistance and high chemical purity. It dries as a thin,
non-corrosive film with excellent adhesion, does not migrate, and is non-freezable. Dow
Corning High Vacuum Grease is a stiff, nonmelting, nonoxidizing, non gumming silicone
lubricating material that is insoluble in most solutions. There are no potential reactions
associated with these lubricants or grease. Other lubricants and greases maybe used as
alternatives to the examples given provided they meet the performance and general
compositional requirements of the nuclear power industry.
2.4.4.2.8
The NAC-STC utilizes seals formed from silicone rubber and Viton. Viton is a silicon
elastomer. Elastomer 0-rings are used for transport cask applications because of their excellent
short-term sealing capabilities, ease of handling, and more economical cost. All of the seal and
gasket materials have stable, non-reactive compositions. There are no potential reactions
associated with the NAC-STC seal materials.
2.4.4.3 General Effects of Identified Reactions
No significant potential galvanic or other reactions have been identified for the NAC-STC. The
only potential chemical reaction identified for the NAC-STC is that of aluminum with the spent
fuel pool water. As discussed in Section 2.4.4.2.2, it is possible at higher temperatures (above
I 50-l 60°F) that a flammable concentration of hydrogen might be generated by the
2.4.4-8
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NAC-STC SAR
Docket No. 71-9235
March 2017
Revision 18
aluminum/water reaction and accumulate beneath the canister shield lid during the canister
closure operations. The danger of potential ignition of the hydrogen is precluded by the
operating controls and procedures presented in Chapter 7. Therefore, no adverse conditions can
result during any phase of cask operations for normal, off-normal, or accident conditions.
2.4.4.4 Adequacy of the Cask Operating Procedures
Based on the results of this evaluation which resulted in only one identified reaction, aluminum
in pool water, it is concluded that the NAC-STC operating controls and procedures presented in
Chapter 7 are adequate to minimize the occurrence of hazardous conditions.
2.4.4.5 Effects of Reaction Products
No significant potential chemical, galvanic, or other reactions have been identified for the NAC
STC. Therefore, the overall integrity of the cask and the structural integrity and retrievability of
the spent fuel is not adversely affected for any cask operations throughout the design basis life of
the cask. Based on the evaluation, there will be no change in the cask or fuel cladding thermal
properties, and there will be no binding of mechanical surfaces, no change in basket clearances,
and no degradation of any safety components either directly or indirectly, since there are no
significant reactions identified .
2.4.4-9
NAC-STC SAR
Docket No. 71-9235, Revision 17C
June 2018
Revision l 8B
Table 2.4-1 Summary of NAC-STC Materials Categories and Operating Environments
ITEM MATERIAL ENVIRONMENT
Stainless Steels/ Alloys 304, 304L, XM-19, 17-4PH, Sealed Internal Ni Alloy, 410 Open Internal/ External
Nonferrous Metals ASTM B 152 Cu, Sealed Internal
606 l -T65 l Aluminum Alloy Open Internal/External
Shielding Materials NS-4-FR, Chemical Copper Enclosed Grade Lead
Criticality Control Materials Boroncarbide Enclosed
Aluminum 1100
Energy Absorbing Materials Balsa Wood, Redwood Enclosed
Cellular Foam/Insulation Silicone (HT-810 & 800), Enclosed Silicone Caulk (Dow Corning)
Lubricants & Greases Never-Seeze® Sealed Internal
Neolube® Open Internal
High Vacuum Grease® by Dow Corning
Seals & Gaskets Silicone Rubber, Viton Sealed Internal
Open Internal/ External
2.4.4-10
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NAC-STC SAR
Docket No. 71-9235, Revision 17B
June 2018
Revision 18B
Bolt Stress Evaluation
The maximum preload on the inner lid bolts in combination with the internal pressure force on
the lid, the 0-ring compression force on the lid, thermally induced loads and the inertial load of
the inner lid and cask contents due to the normal conditions of transport I-foot drop must not
exceed the allowable strength of the inner lid bolts.
A complete range of impact orientations is evaluated, from an end impact at 0° to a side impact at
90°, using 5° increments. A design acceleration of 20g is considered for all impact orientations
in the bolt evaluation for normal conditions for both impact limiter designs (redwood and balsa).
The cask contents weight for the CY-MPC is used since it bounds the cask content weight of the
Yankee-MPC configuration. The details of this evaluation are described and an example
calculation is provided in Section 2.10.8.2 for the hypothetical accident condition. Normal
conditions of transport results are summarized in Tables 2.6.7.5-1 and 2.6.7.5-2, corresponding
to a "hot" condition and a "cold" condition, respectively. The hot condition bolt temperatures are
assumed to be 200°F, as summarized in Table 3.4-5. The cold condition bolt temperature is
assumed to be -20°F, in accordance with regulatory requirements. Physical properties for the
SB-637 Grade N07718 nickel alloy bolts are conservatively taken at 270°F for both hot and cold
conditions. As defined in Table 2.1.2-1, the allowable maximum bolt stress for normal
conditions for primary membrane stress is two times the design stress intensity, 2Sm, resulting in
an allowable direct tension stress of 94.5 ksi at 270°F. As shown in Tables 2.6.7.5-1 and 2.6.7.5-
2, the total bolt stress is calculated to be less than the allowable stress for normal conditions of
transport. The minimum margin of safety is +0.06.
Bolt Thread Engagement Evaluation
The ultimate load capacity of the inner lid bolt/top forging threaded connection relative to the
ultimate tensile load capacity of the inner lid bolt is evaluated to ensure that the length of
engagement is sufficient to develop the full strength of the bolt. The inner lid bolt holes have
threaded inserts to protect the threads during the installation and removal of the bolts.
Component Description
Inner lid Bolt 11/2-8UN
SB-637, Grade N07718 Nickel Alloy Steel Bolting Material
Length in cask body= 9.75 inches
Su= 174.7 ksi at 270°F
2.6.7.5-5
NAC-STC SAR
Docket No. 71-9235, Revision 17B
Threaded Insert Helicoil #4190-24 CN x 2.50
(AMS 7245) 18-8 Stainless Steel
Length of insert= 2.50 inches
O.D. = I 5/8 - 8 UN Thread
Su= 200.0 ksi
Top Forging (Cask Body) Type 304 Stainless Steel
Thread depth= 3.0 - 0.125 = 2.875 in
Su = 62.9 ksi at 270°F
Bolt Strength
Tensile Area, At
Tensile Strength, Su
Bolt-Tensile Load Capacity, PBLC
= 1.492 in2 (I 1/2 - 8 UN Thread)
= 174.7 ksi
(1.492)(174,700)
260,650 lbs
Threaded Insert/Bolt Interface
Thread Size = 1 l/2-8UN
Engaged Length = Lbolt - hvasher - tiid - dset-down = 2.20 in
where:
Lbolt Length of bolt= 9.75 in
twasher Thickness of washer= 0.315 in [8 mm]
tiid = Thickness of lid at bolt location= 7.10 in dset-down Insert set-down distance= 0.14 in
External (Bolt) Thread Shear Area
where:
ASs = (n)(n)(Le)(Knmax)[(l /211) + (0.57735)(Esmin - Knmax)f
= 5.660 in2
n = 8 threads/in
Le = 2.20 in
Esmin = I .4093 in Min. Pitch Diameter of External Threads
Knmax = 1.390 in Max. Minor Diameter ofJnternal Threads
* FED-STD-H28 (1963), Page 103.
2.6.7.5-6
June 2018
Revision 18B •
•
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NAC-STC SAR
Docket No. 71-9235, Revision l 7B
Bolt Thread-Tensile Load Capacity, PBT
Internal (Insert) Thread Shear Area
= (5.660)(0.5* X 174,700)
= 494,400 lbs
June2018
Revision l 8B
ASn = (n)(n)(Le)(Dsmin)[(l/2n) + (0.57735)(Dsmin - Enmax)]**
= 7.704 in2
where:
Dsmin = 1.4828 in Min. Major Diameter of External Thread
Enmax = 1.4283 in Max. Pitch Diameter of Internal Thread
Insert Thread-Tensile Load Capacity, Pm
Threaded Insert/Top Forging Interface
Thread Size = I 5/8 - 8 UN
Engaged Length
Insert: Su
Top Forging: Su
= 2.20 in
= 200 ksi
= 62.9 ksi
External (Insert) Thread Shear Area
= (7.704) (0.5* X 200,000)
= 770,400 lbs
ASs = (rc)(n)(Le)(Knmax)[(]/2n) + (0.57735)(Esmin - Knmax)]**
= 6.164 in2
* Shear Strength Conservatively Assumed= (0.5)(Tensile Strength).
** FED-STD-H28 ( 1963), page 103 .
2.6.7.5-7
NAC-STC SAR
Docket No. 71-9235, Revision 17B
where:
n = 8 threads/inch
Le = 2.20 in
Esmin = 1.5342 in Min. Pitch Diameter of External Threads
Knmax = 1.515 in Max. Minor Diameter of Internal Threads
Insert Thread-Tensile Load Capacity, Pno = (6.164)(0.5* x 200,000)
= 616,400 lbs
Internal (Top Forging) Thread Shear Area
June 2018
Revision 18B
ASn = (n)(n)(Le)(Dsmin)[(l/2n) + (0.57735)(Dsmin - Enmax)]**
= 8.343 in2
where:
Dsmin = 1.6078 in Min. Major Diameter of External Thread
Enmax = 1.5535 in Max. Pitch Diameter oflnternal Thread
Top Forging Thread - Tensile Load Capacity, PTFT = (8.343)(0.5* x 62,900)
= 262,390 lbs
Component
Inner Lid Bolt
Bolt Thread
Insert I.D. Thread
Insert O.D. Thread
Top Forging Thread
Ultimate Load Capacity
(lbs)
260,6,50
494,400
770,400
616,400
262,390
Since the mm1mum Tensile Load Capacity of the threaded joint (262,390 lbs) exceeds the
maximum Tensile Load Capacity of the inner lid bolt (260,650 lbs), the load capacity of the
inner lid bolt is the controlling load capacity of the joint strength, and the design requirements
are satisfied. The inner lid bolt threaded-joint design is satisfactory.
* Shear Strength Conservatively Assumed = (0.5)(Tensile Strength).
** FED-STD-H28 (1963), page 103.
2.6.7.5-8
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NAC-STC SAR
Docket No. 71-9235, Revision 17B
2.6.11 Fabrication Conditions
June 2018
Revision 18B
The process of manufacturing the NAC-STC can introduce thermal stresses in the inner and outer
shells as a result of pouring molten lead between them. These thermal stresses are evaluated in
this section to provide assurance that the manufacturing process does not adversely affect the normal
operation of the cask or its ability to survive an accident. Any residual stresses in the containment
vessel shell due to inelastic strain associated with the secondary local bending stresses, which result
from the lead pour thermal gradient, must be considered in the total stress range for normal and
accident load conditions according to Regulatory Position 7 of Regulatory Guide 7.6. Residual
stresses in the containment vessel and the outer shell induced by shrinkage of the lead shielding
after the lead pouring operation are relieved early in the life of the cask because of the low creep
strength of lead.
For the lead pouring process, the temperatures of the cask shells are controlled between 640°F
(338°C) and 740°F (393°C), and the maximum lead temperature before pouring is 790°F (421 °C).
Heating of the cask is performed using heaters inside the inner shell and heating rings around the
outside of the outer shell. Heat up is time controlled, consistent with maintaining shell
temperatures uniformly. The heating procedures ensure that the surface temperature of the cask
does not exceed 800°F (427°C). The shell temperatures are measured by thermocouples attached
to the shell surfaces. A portable thermometer is also used to measure temperature at any location.
Cask heating is carried out after all of the preparations have been completed (including melting of
the lead) in order to minimize the time that the cask is at elevated temperatures.
The lead is poured after the cask reaches the specified temperatures. Prior to lead pouring, the cask
flange area is heated with hand-held burners to between 640°F (338°C) and 740°F (393°C).
Pouring is carried out continuously using a filling tube with its open end maintained under the lead
surface. The pouring time is kept as short as possible. During pouring, the interior heaters and
exterior heating rings are continuously energized.
The cooling process consists of sequentially turning the exterior heating rings and interior heaters
off, starting from the lowest point, and of spraying the cask with water from the outside. A layer
of molten lead is maintained until the upper surface starts to solidify. This process allows the molten
lead to fill the open space below it created by the lead shrinkage as it cools.
The basic requirements and procedures for the NAC-STC lead pour operations are described in
Section 8.4.2 and 8.4.3 .
2.6.11-1
•
THIS PAGE INTENTIONALLY LEFT BLANK •
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NAC-STC SAR
Docket No. 71-9235, Revision l 7B
2.6.11.1 Lead Pour
2.6.11.1.1 Cask Shell Geometry
At 70°F, the Type 304 stainless steel shell geometry is as follows:
Inner Shell
Inside Diameter (di)
Outside Diameter (do)
Shell Thickness (ti)
Outer Shell
Inside Diameter (Di)
Outside Diameter (Do)
Shell Thickness (To)
= 71.0 in
= 74.0 in
= 1.5 in
= 81.4 in
= 86.7 in
= 2.65 in
2.6.11.1.2 Stresses Due to Lead Pour
June 2018
Revision l 8B
As stated in the Lead Pour Procedures in Sections 8.4.2 and 8.4.3, the maximum lead temperature
during the pouring operations is 790°F. Assuming that the lead and the inner and outer shells are
uniformly at 790°F, the hydrostatic pressure produced by the column of lead is:
p = ph
= 66 psi
where
p = 0.41 lb/in3 (lead density)
h = 161 in (maximum height of lead column)
2.6.11.1-1
NAC-STC SAR
Docket No. 71-9235, Revision l 7B
At 790°F, the shell geometric dimensions are:
where
d~ = do ( 1 + a ~ T)
n; =Di(l+a~T)
t' = t (1 + a ~ T)
a = J 0.09 x 1 o-6 in/in/°F at 790°F (stainless steel)
~T = 790 - 70 = 720°F
d~ = (74.0) [1 + (1 o.o9 x I o-6)(720)]
= 74.54 in
d~ = [71.o;
74·0
] [I+ (10.09 x 10-6)(720)]
= 73.03 in
n; = (81.4)[1 + (10.09 X 10-6)(720)]
= 81.99 in
, [81.4 + 86.7] Dm= 2
[(1+(10.09xl0-6)(720)]
= 84.66 in
2.6.11.1-2
June 2018
Revision 18B
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NAC-STC SAR
Docket No. 71-9235, Revision 17B
' ti = (1.50) [I+ (10.99 x 10-6)(720)]
= 1.51 in
' t 0 = (2.65) [1 + (I 0.99 X 1 o-6)(720)]
= 2.669 in
June 2018
Revision 18B
The inner shell is subjected to an external hydrostatic pressure, and the outer shell to an internal
hydrostatic pressure, of 66 psi. This causes the inner shell to decrease in diameter and the outer
shell to increase in diameter.
The inner shell decreases in size radially (Roark and Young, 5th ed., Case 1 b, page 448):
where
b.rm
(-66)(74.54/2)2 - 0.00251 in (24.17 x10 6 )(1.51)
(-66)(73.03/2)2 - 0.00241 in (24.17 xl0 6 )(1.51)
E = 24.17 x 106 psi at 790°F.
The outer shell increases in size radially:
q(D{,,/2) 2
Et0
(66
)(8
1.9912)
2 = 0.00172 in (24.17 x106 )(2.669)
(66)(84.66/2)2 = 0.00183 in (24.17 x10G )(2.669)
2.6.11.1-3
NAC-STC SAR
Docket No. 71-9235, Revision 17B
The shell geometries at 790°F and 66 psi hydrostatic pressure are:
d"o = 74.54 - (2)(0.00251) = 74.535 in
D"i = 81.99 + (2)(0.00172) = 81.993 in
d\n = 73 .03 - (2)(0.00241) = 73 .025 in
D"m = 84.66 + (2)(0.00183) = 84.666 in
June 2018
Revision 18B
The hoop stresses are evaluated at the mean diameter of the inner and outer shells at 790°F:
= Pd;~ 2t'
I
C-66)C73·025
) = -1596 psi (inner shell) (2)(1.51)
= PD;·n
2t~
C66)CB 4·55
) = 104 7 psi ( outer shell) (2)(2.669)
These stresses are negligible.
2.6.11.1-4
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NAC-STC SAR
Docket No. 71-9235, Revision I 7B
2.6.11.2 Cooldown
2.6.11.2.1 Hoop Stresses
June 2018
Revision I 8B
Lead decreases in volume during solidification. As the lower lead region solidifies, the molten lead
above fills the shrinkage void between the solidifying lead and the inner and outer shells, thus,
maintaining the 66 psi pressure on the shells.
The stress-free inner and outer radii of the solidified lead can be calculated (Roark and Young, 5th
ed., Cases I a and 1 c, page 504) as:
L'ia - q [ 2ab2 l qa [a 2 - b2 - v] = Eq a(v - I) E a 2 -b2 E a 2 -b2
=-0.001104 in
L'ib + v - - = - b(v - I) = q b [ a 2
+ b 2 l q [ 2a
2 b l q
E a 2 -b2 E a 2 -b2 E
= -0.001004 in
where
q = 66 psi (pressure)
E = 1.47 x 106 psi at 620°F (modulus of elasticity)
v = 0.4 (Poisson's ratio)
a = D"i/2 = 81.993/2 = 40.9965
b = d"o/2 = 74.535/2 = 37.2675
then
Roi = 40.9965 - 0.001104 = 40.9954
2.6.1 1.2-1
NAC-STC SAR
Docket No. 71-9235, Revision 17B
Ri1 = 37.2675 - 0.001004 = 37.2665
When cooled to 70°F, the inside radius of the lead is such that:
where
R;e = inside radius of the stress-free lead at 70°F
a = 20.4 x 1 o-6 in/in/°F
!-.T = 550°F (620-70)
then
likewise
_R_oe __ R' 1 + a!-.T - oe
R~e = 40.5405 in
June 2018
Revision 18B
The outside radius of the stress-free inner shell is 74.0/2 = 37.0 inches, which is larger than the
stress-free inner radius of the lead shell. Therefore, there exists an interface pressure between the
lead and the inner shell after cooling to 70DF. The interface pressure, when acting on the lead
cylinder and inner shell, is such that the inner radius of the lead cylinder is the same as the outer
radius of the inner shell (Roark and Young, 5th ed., Case 1 a, page 504).
2.6.11.2-2
•
•
•
•
•
•
NAC-STC SAR
Docket No. 71-9235, Revision 17B
where
then
= b + q b [ a 2 + b 2 + v] E a 2
- b2
R;e = inside radius of lead cylinder at 70°F
V = 0.4
E = 2.28 x I 06 psi at 70°F
a = 40.5405 in
b = 36.853 in
= 36.853 + ( 36.853q ) ((40.5405)2
+ (36.853)2 + 0.4)
2.28 x10 6 (40.5405) 2 - (36.853) 2
= 36.853 + 1.765 x I o-4q
June 2018
Revision 18B
The outside radius of the inner shell at 70°F under the interface pressure, q, (Roark and Young,
5th ed., Case le, page 504) is:
where
ro = as -L'ias
= as _ qa s (a; + b; _ vJ E a 2
- b2 s s
ro = outside radius of inner shell at 70°F
as = 74.0/2 = 37.0 in
bs = 71.0/2 = 35.5 in
2.6.11.2-3
NAC-STC SAR
Docket No. 71-9235, Revision l 7B
then
E = 28.3 x 106 psi at 70°F
V = 0.275
= _ ( 37.0q :( (37.0)2
+ (35.5)2
_ O 275] ro 37.0 6 2 2 · 28.3 X 10 (37.0) - (35.5)
= 37.0 - 3.)25 X 10-5q
Equating R;e and ro and solving for q:
q = 708 psi interface pressure
The lead shell geometry is:
((36.853)(708)) ((40.5405)
2 + (36.853)2
- 0.4) 2.28 Xl06 (40.5405)2 - (36.853) 2
= 36.969 in
= 40.5405 + ( (708) ) ( (2)( 40.5405)(36.853)2
) 2.28 Xl0 6 (40.5405) 2 - (36.853) 2
= 40.66 in
June 2018
Revision l 8B
The interference between the lead shell and the inner shell is 0.142 inch (37.0 - 36.853). To fully
accommodate this interference, the lead must undergo a strain of 0.172/36.853 = 0.004 or 0.4
percent. From Figure 24 ofNUREG/CR-0481, the lead stress for the above strain is 800 psi. The
corresponding interface pressure for this stress in the lead shell is:
2.6.11.2-4
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•
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NAC-STC SAR
Docket No. 71-9235, Revision l 7B
q [
, 2 , 2 J = (S) Roe - Ric , 2 , 2
Roe + Ric
= (800) ((40.660)2
- (39.969)2
) ( 40.660)2 + (39.969)2
= 76 psi interface pressure
The change in geometry of the inner shell for this interface pressure is:
L1a [ -76 ] [(2)(37.0)(35.5)2
] - 28.3 X 106 (37.0)2 - (35.5) 2
= 0.0023 in
This can conservatively be neglected in the analysis. The inner shell hoop stress is:
Shis = (-76) [(37.0)2 + (35.5)21 (37.0) 2 - (35.5)2
=-1837psi
This stress is negligible.
2.6.11.2.2 Axial Stresses
June 2018
Revision l 8B
Axial stresses also develop in the lead shell and inner shell during fabrication as a result of the
unequal shrinkage of the lead and steel shells. Assume bonding of the lead shell to the inner shell
during the cool down process after completion of lead pouring. The strain in the lead, when cooled
to 70°F, is:
Ee = (a.1 - Us)L1T
= 0.0060 in/in or 0.60 percent
where
2.6.11.2-5
NAC-STC SAR
Docket No. 71-9235, Revision 17B
a.e = 20.4 X 1 o-6 in/in/°F
a5 = 9.56 x 10-6 in/in/°F
LiT = 620 - 70 = 550°F
June 2018
Revision 18B
Extrapolating from Figure 24 ofNUREG/CR-0481 for this strain, an axial stress of approximately
825 psi exists in the lead shell. The total force in the lead caused by assuming no deformation of
the inner shell is:
PsPb = Pe Ae
= 825n[(40.7)2 - (37.0)2]
= 745,120 lb tensile force
The corresponding compression stress in the inner shell to maintain equilibrium is:
-745,120
n[(37.0) 2 - (35.5) 2
]
= -2180 psi
This stress is negligible.
This is a highly conservative estimate of the compressive stress that can develop in the inner shell
for the following reasons:
1. It assumes no axial deformation of the inner shell and no load development in the
outer shell.
2.6.11.2-6
•
•
•
• NAC-STC SAR
Docket No. 71-9235, Revision 17B
June2018
Revision 18B
2. Creep in the lead is neglected. This also reduces the stress and force in the_ lead
(Section 2.6.11.3).
3. It assumes the strain is uniform through the thickness of the lead shell, i.e., no shear
strain exists in the plane formed by the radial axis and the longitudinal axis. A
particle away from the inner shell should develop Jess strain, consequently lower
stress, than a particle adjacent to the inner shell; this also reduces the total force in
the lead shell.
2.6.11.2.3 Effects of Temperature Differential During Cooldown
The preceding analyses assume that the inner and outer shells and the lead are always at the same
temperature at any time during the cooldown process. This assumption may not be true under
actual conditions. However, because of the high thermal conductivity of the stainless steel and the
lead and because of the time-controlled cooldown process, the temperature differential between
any two of the above shells is kept to a minimum.
• If the inner shell is cooler than the lead, the interference between them as well as the corresponding
interface pressure and hoop stresses are Jess than for the case of equal temperatures. Hence, the
preceding analysis is conservative.
•
If the inner shell is hotter than the lead shell, an analysis is required. As described in the Lead Pour
Procedures in Section 8.4, the maximum allowed temperature differential measured between the inside
surface of inner shell and the outside surface of outer shell is 100°F (Section 8.4.2) and 160°F (Section
8.4.3) for the alternate lead pour procedure. Considering the bounding temperature differential of 160°F
for the inside surface of inner shell and the outside surface of outer shell, the temperature differential
between inner shell and lead shell is calculated to be less than J 00°F. Conservatively using a 100°F
temperature differential for the inner shell and the lead shell by assuming the temperature of the inner
shell to be 170°F and that of the lead to be 70°F, the inner radius of the stress-free lead shell at
70°F is 36.853 inches ( R;e ); the outer radius of the inner shell at 170°F is:
R =37.0 [1 +(8.71 x 10-6)(100)]
= 37.032 in
2.6.11.2-7
NAC-STC SAR
Docket No. 71-9235, Revision 17B
June 2018
Revision l 8B
The interference between the inner shell and the lead is 37.032 - 36.853 = 0.179 inch. To fully
accommodate this interference, the lead has to undergo a strain of 0.179/36.853 = 0.0049 inch/inch
or 0.49 percent. From Figure 24 ofNUREG/CR-048 I, the hoop stress in the lead is approximately
810 psi for a 0.0049 inch/inch strain. The interface pressure is:
q = (810) [( 40.5405)2
- (36.853)2
]
(40.5405) 2 + (36.853) 2
= 77 psi
The hoop stress in the inner shell becomes:
[(37)
2 + (35.5)2 l
Shis = (-77) (37)2 - (35.5)2
= -1862 psi
Note that the thermal expansion or contraction of a shell subjected to a constant pressure does not
affect the hoop stress; i.e.,
Sh = [(ka)2 + (kb)2 ]= [a
2
+ b2
]
q (ka)2 - (kb)2 q a 2 - b2
where
k = 1 + a ~T
This -1862 psi hoop stress (the inner shell is assumed to be 100°F hotter than the lead shell) reduces
to the previously calculated hoop stress of -1837 psi as both the inner shell and lead reach an
ambient temperature of 70°F. This does not take into account the beneficial effect of the creep
properties of the lead.
The axial stress in the inner shell also increases when the inner shell is 100°F hotter than the lead
shell. The axial stress of -2180 psi calculated when both the inner shell and lead shell are at 70°F
is recalculated for the inner shell temperature of l 70°F, a= 8.71 x 10-6 inch/inch/°F (Type 304
stainless steel):
2.6.11.2-8
•
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NAC-STC SAR
Docket No. 71-9235, Revision l7B
Se = (20.38 - 9.56)(10-6)(620 - 70) + (8.71 X 10-6)(] 70 - 70)
= 0.00595 in/in or 0.595 percent
June 2018
Revision l 8B
Referring to Figure 24 of NUREG/CR-048 I, the axial stress in the lead is approximately 820 psi.
The corresponding axial stress in the inner shell is -2 I 67 psi. As before, cooling of the inner shell
reduces this stress. The previous assumptions apply in arriving at this inner shell compressive
stress.
Temperature differentials between the inner and outer shells are of no consequence, since the axial
restraint between them is welded in place after cooldown, when the cask is at a uniform ambient
temperature. Welding of the outer shell and the bottom inner forging to the bottom outer forging
after cooldown is, therefore, a necessary fabrication step.
The question of buckling of the inner shell due to the combined effect of external pressure and
fabrication inaccuracies must also be addressed. According to the "ASME Boiler and Pressure
Vessel Code," Article NE-4221. I, the difference between the maximum and minimum inside
diameters at any cross section shall not exceed I percent of the nominal diameter at the cross
section under consideration. This amounts to (0.01)(71.0) = 0.71 inch. The relation between the
initial radial deviation, ro1, and the maximum and minimum diameter (Timoshenko and Gere,
Figure 7-10) is:
Dmax = Dnom + 2ro I
Dmin = Dnom - 2ro 1
thus
Dmax - Dmin = 4ro I
or
~D = 4ro1
Hence, the maximum initial radial deviation allowed is:
ffimax = ~D/4 = 0.71/4 = 0.1775 in
2.6. I 1.2-9
NAC-STC SAR
Docket No. 71-9235, Revision l7B
From Timoshenko and Gere, equation (7-15), page 2 93:
where
= 12,856 psi
E =27.76xl06 psiat170°F
V = 0.275
h = shell thickness= 1.50 in
R = shell radius= 36.25 in
Then from Timoshenko and Gere, equation (7-12), page 289:
qoc ~ [ 4 (I~ v 2)J[;r = Ser [: J
= 532 psi
June 2018
Revision l 8B
When the cylinder has fabrication inaccuracies, the external pressure, qYr, required to produce
yielding in the extreme fibers can be solved in the following equation (Timoshenko and Gere,
equation (e), page 296):
where
2 [Syr J Syr q YP - --;;;- + (I + 6mn) qcr g YP + --;;;-qcr = 0
Syp= 26,150 psi at l 70°F for Type 304 stainless steel
m = R/h = 36.25/1 .50 == 24.1667
n = mi/R = mi/36.25
2.6.11.2-10
•
•
•
•
•
•
NAC-STC SAR
Docket No. 71-9235, Revision 17B
then
qvP2 - [l 082 + (I + 4m1)(532)] qvP + 575,660 = 0
June 2018
Revision 18B
The value of m1 can vary from 0.0 inches (perfect cylinder) up to 0.1775 inch (maximum allowed
according to the "ASME Boiler and Pressure Vessel Code"). Solving qyp for varying values of m1
gives the following:
Initial Radial Yield
Deviation Pressure
m1 (in) gyp (psi)
0.001 530
0.06 443
0.12 389
0.1775 351
Thus, the margin of safety against yielding for the inner shell with maximum allowed radial
deviation subjected to 77 psi lead pressure (inner shell temperature is assumed to be 100°F higher
than lead temperature) is:
351 MS = - - 1 = +3.55
77
Since the margin of safety for this conservative load case is positive, the inner shell does not buckle
when subjected to the lead pressure produced during the cooling of the cask .
2.6.11.2-11
•
THIS PAGE INTENTIONALLY LEFT BLANK •
•
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• ANAC fnfl INTE RNA TIO NAL
Atlanta Corporate Headquarters: 3950 East JOfleS Bridge Road, Norcross, Georgia 30092 USA Phone 770-447-1144, Fax 770-447-1797, www.nacintl.com
June 2018
Revision 18B
•
•
•
NAC-STC SAR Docket No. 71-9235
June 2018 Revision 18B
List of Effective Pages
Chapter 1 Page 2.4-1 ..................................... Revision 18
Page 2.4.1-1 .................................. Revision 18
Page 1-i thru 1-iv .......................... Revision 18 Page 2.4.2-1 .................................. Revision 18
Page 1-v ...................................... Revision 18B Page 2.4.3-1 .................................. Revision 18
Page 1-vi thru I-ix ........................ Revision 18 Pages 2.4.4-1 thru 2.4.4-6 ............. Revision 18
Pages 1-1 thru 1-12 ....................... Revision 18 Pages 2.4.4-7 thru 2.4.4-8 ........... Revision 18B
Pages 1.1-1 thru 1.1-3 ................... Revision 18 Page 2.4.4-9 .................................. Revision 18
Page 1.1-4 ................................... Revision 18B Pages 2.4.4-10 ............................. Revision 18B
Pages 1.1-5 thru 1.1-46 ................. Revision 18 Page 2.4.5-1 .................................. Revision 18
Pages 1.2-1 thru 1.2-8 ................... Revision 18 Page 2.4.6-1 .................................. Revision 18
Page 1.2-9 ................................... Revision 18B Pages 2.5.1-1 thru 2.5.1-3 8 ........... Revision 18
Pages 1.2-10 thru 1.2-42 ............... Revision 18 Pages 2.5.2-1 thru 2.5.2-29 ........... Revision 18
Pages 1.2-43 thru 1.2-44 ............. Revision 18B Pages 2.6-1 thru 2.6-2 ................... Revision 18
Pages 1.2-45 thru 1.2-49 ............... Revision 1 8 Pages 2.6.1-1 thru 2.6.1-7 ............. Revision 18
Page 1.3-1 ..................................... Revision 18 Pages 2.6.2-1 thru 2.6.2-8 ............. Revision 18
Pages 1.4-1 thru 1.4-24 ................. Revision 18 Page 2.6.3-1 .................................. Revision 18
Page 2.6.4-1 .................................. Revision 18
Chapter 2 Pages 2.6.5-1 thru 2.6.5-2 ............. Revision 18
Page 2.6.6-1 .................................. Revision 18
Pages 2-i thru 2-lxviii .................... Revision 18 Page 2.6.7-1 .................................. Revision 18
Page 2-1 ........................................ Revision 18 Pages 2.6.7.1-1 thru 2.6.7.1-17 .... Revision 18
Pages 2.1.1-1 thru 2.1.1-2 ............. Revision 18 Pages 2.6.7.2-1 thru 2.6.7.2-19 ..... Revision 18
Pages 2.1.1-3 thru 2.1.1-4 ........... Revision 18B Pages 2.6.7.3-1 thru 2.6.7.3-11 ..... Revision 18
Pages 2.1.1-5 ................................. Revision 18 Pages 2.6.7.4-1 thru 2.6.7.4-59 ..... Revision 18
Pages 2.1.2-1 thru 2.1.2-5 ............. Revision 18 Pages 2.6.7.5-1 thru 2.6.7.5-4 ....... Revision 18
Pages 2.1.3-1 thru 2.1.3-15 ........... Revision 18 Pages 2.6.7.5-5 thru 2.6.7.5-8 ..... Revision 18B
Pages 2.2-1 thru 2.2-8 ................... Revision 18 Pages 2.6.7.5-9 thru 2.6.7.5-13 ..... Revision 18
Pages 2.3.1-1 thru 2.3.1-2 ............. Revision 18 Pages 2.6.7.6-1 thru 2.6.7.6-13 ..... Revision 18
Pages 2.3.2-1 ................................. Revision 18 Pages 2.6.7.7-1 thru 2.6.7.7-5 ....... Revision 18
Pages 2.3.2-2 thru 2.3.2-4 ........... Revision 18B Page 2.6.8-1 .................................. Revision 18
Pages 2.3.2-5 ................................. Revision 18 Page 2.6.9-1 .................................. Revision 18
Pages 2.3.3-1 thru 2.3.3-2 ............. Revision 18 Page 2.6.10-1 ................................ Revision 18
Pages 2.3.4-1 thru 2.3.4-3 ............. Revision 18 Pages 2.6.10.1-1 thru
Pages 2.3.5-1 thru 2.3.5-2 ............. Revision 18 2.6.10.1-2 ................................ Revision 18
Pages 2.3.6-1 thru 2.3.6-5 ............. Revision 18 Pages 2.6.10.2-1 thru
Page 2.3.7-1 .................................. Revision 18 2.6.10.2-4 ................................ Revision 18
Page 2.3.8-1 .................................. Revision 18
I of 7
NAC-STC SAR Docket No. 71-9235
June2018 Revision 18B
List of Effective Pages (continued)
Pages 2.6.10.3-1 thru Pages 2.6.13.2-1 thru
2.6.10.3-7 ................................. Revision 18 2.6.13.2-7 ................................. Revision 18
Page 2.6.11-1 .............................. Revision 18B Pages 2.6.13 .3-1 thru
Pages 2.6.11.1-1 thru 2.6.13.3-4 ................................. Revision 18
2.6.11.1-4 .............................. Revision 18B Pages 2.6.13 .4-1 thru
Pages 2.6.11.2-1 thru 2.6.13.4-5 ................................. Revision 18
2.6.11.2-11 ............................ Revision 18B Pages 2.6.13.5-1 thru
Page 2.6.11.3-1 .............................. Revision 18 2.6.13.5-2 ................................. Revision 18
Pages 2.6.12-1 thru Pages 2.6.13.6-1 thru
2.6.12-5 .................................... Revision 18 2.6.13 .6-2 ................................. Revision 18
Page 2.6.12.1-1 .............................. Revision 18 Pages 2.6.13.7-1 thru
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2.6.12.3-7 ................................. Revision 18 Page 2.6.13.10-1 ............................ Revision 18
Pages 2.6.12.4-1 thru Pages 2.6.13.11-1 thru
2.6.12.4-3 ................................. Revision 18 2.6.13.11-3 ............................... Revision 18
Pages 2.6.12.5-1 thru Pages 2.6.13.12-1 thru
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Pages 2.6.12.6-1 thru Pages 2.6.14-1 thru
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Pages 2.6.12.8-1 thru Pages 2.6.14.2-1 thru
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Pages 2.6.12.9-1 thru Pages 2.6. I 4.3-1 thru
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Page 2.6.14.9-1 .............................. Revision 18
2 of7
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NAC-STC SAR Docket No. 71-9235
June 2018 Revision l 8B
List of Effective Pages (continued)
Page 2.6.14.10-1 ........................... Revision 18 Pages 2.6.16.6-1 thru
Pages 2.6.14.11-1 thru 2.6.16.6-3 ................................ Revision 18
2.6.14.11-5 .............................. Revision 18 Pages2.6.16.7-l thru
Pages 2.6.14.12-1 thru 2.6.16.7-12 .............................. Revision 18
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Page 2.6.15-1 ................................ Revision 18 2.6.16.8-7 ................................ Revision 18
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2.6.15.1-2 ................................ Revision 18 Page 2.6. 16.10-1 ........................... Revision 18
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2.6.15.2-7 ................................ Revision 18 2.6.16.11-4 .............................. Revision 18
Pages 2.6.15 .3-1 thru Pages 2.6.16.12-1 thru
2.6.15.3-4 ................................ Revision 18 2.6.16.12-2 .............................. Revision 18
Pages 2.6.15.4-1 thru Pages 2.6.16.13-1 thru
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Page 2.6.15.5-1 ............................. Revision 18 Page 2.6.16.14-1 ........................... Revision 18
Pages 2.6.15 .6-1 thru Pages 2.6.17-1 thru
2.6.15.6-3 ................................ Revision 18 2.6.17-13 ................................. Revision 18
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Pages 2.6.15.11-1 thru Pages 2.6.20-1 thru
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Pages 2.6.15.12-1 thru Pages 2.6.21-1 thru
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Pages 2.6.16-1 thru Pages 2.7-1 thru 2.7-2 ................... Revision 18
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Pages 2.6.16.2-1 thru Pages 2.7.1.2-1 thru
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Pages 2.6.16.3-1 thru Pages 2.7.1.3-1 thru
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Pages 2.6.16.4-1 thru Pages 2. 7 .1.4-1 thru
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Pages 2.6.16.5-1 thru Pages 2.7.1.5-1 thru
2.6.15.5-3 ................................ Revision 18 2.7.1.5-3 .................................. Revision 18
3 of7
L_
NAC-STC SAR Docket No. 71-9235
June 2018 Revision 18B
List of Effective Pages (continued)
Pages 2.7.1.6-1 thru
2.7.l.6-16 ................................. Revision 18
Page 2.7.2-1 ................................... Revision 18
Pages 2.7.2.1-1 thru
2.7.2.1-5 ................................... Revision 18
Pages 2.7.2.2-1 thru
2.7.2.2-9 ................................... Revision 18
Pages 2.7.2.3-1 thru
2.7.2.3-6 ................................... Revision 18
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List of Effective Pages ( continued)
4 drawings in Section 2.10.12 Chapter 3
Pages 2.11.1-1 thru 2.11.1-2 ......... Revision 18 Page 3-i ....................................... Revision 18B
Pages 2.11.2-1 thru 2.11.2-2 ......... Revision 18 Pages 3-ii thru 3-iii ........................ Revision 18
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June 2018 Revision l SB
List of Effective Pages (continued)
Chapter 4 Pages 5.7.6-1 thru 5.7.6-22 ............ Revision 18
Page 5.8-1 ...................................... Revision 18
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Chapter 6
Chapter 5
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NAC-STC SAR Docket No. 71-9235
List of Effective Pages (continued)
Page 6.9 .1-1 .................................. Revision 18
Page 6.9 .2-1 .................................. Revision 18
Chapter 7
Pages 7-i thru 7-ii .......................... Revision 18
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Chapter 8
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Page 8-1 ........................................ Revision 18
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Chapter 9
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NAC-STC SAR
Docket No. 71-9235, Revision 17B
Table of Contents
June 2018
Revision 18B
3.0 THERMAL EVALUATION .......................................................................................... 3.1-l
3.1 Discussion ......................................................................................................................... 3.1-1
3.1.1 Directly Loaded (Uncanistered) Fuel. .............................................................. 3.1-'3
3.1.2 Canistered Yankee Class Fuel ......................................................................... 3.1-4
3.1.3 Canistered Connecticut Yankee Fuel ............................................................... 3.1-6
3.1.4 Canistered Greater Than Class C Waste .......................................................... 3.1-8
3.1.5 Directly Loaded (Uncanistered) PWR High Burnup Fuel ............................... 3.1.9
3.2 Summary of Thermal Properties of Materials .................................................................. 3.2-l
3.2.1 Conductive Properties ...................................................................................... 3.2-1
3.2.2 Radiative Properties ......................................................................................... 3.2-1
3 .2 .3 Convective Properties ...................................................................................... 3 .2-7
3.2.4 Neutron Shield (NS-4-FR) Thermal Conductivity .......................................... 3.2-8
3.3 Technical Specifications for Components ........................................................................ 3.3-l
3.3.1 Radiation Protection Components ................................................................... 3.3-1
3.3.2 Safe Operating Ranges ..................................................................................... 3.3-2
3.4 Thermal Evaluation for Normal Conditions of Transport ................................................ 3.4-1
3.4.1 Thermal Models ............................................................................................... 3.4-1
3.4.2 Maximum Temperatures ................................................................................ 3.4-26
3.4.3 Minimum Temperatures ................................................................................. 3.4-30
3.4.4 Maximum Internal Pressure ........................................................................... 3.4-30
3.4.5 Maximum Thermal Stresses .......................................................................... 3.4-44
3.4.6 Summary ofNAC-STC Performance for Normal Conditions of Transport .. 3.4-44
3.4.7 Normal Heat-up Transient ............................................................................. 3.4-44
3-i
NAC-STC SAR
Docket No. 71-9235
Table of Contents
(Continued)
March 2017
Revision 18
3.5 Hypothetical Accident Thermal Evaluation ...................................................................... 3.5-l
3.5.1 Thermal Model. ................................................................................................ 3.5-I
3.5.2 Package Conditions and Environment.. ........................................................... 3.5-4
3.5.3 Package Temperatures ..................................................................................... 3.5-5
3.5.4 Maximum Internal Pressure ............................................................................. 3.5-7
3.5.5 Maximum Thermal Stresses .......................................................................... 3.5-10
3.5.6 Evaluation of Package Performance for Hypothetical Accident
Thermal Conditions ....................................................................................... 3.5-10
3.6 Thermal Evaluation - STC-LACBWR ......................................................................... 3.6-1
3.6.1 Discussion-STC-LACBWR .............................................................................. 3.6.1-1
3.6.2 Summary of Thermal Properties of Materials- STC-LACBWR ...................... 3.6.2-1
3.6.3 Technical Specifications for Components - STC-LACBWR ....................... 3.6.3-1
•
3 .6.3 .1 Radiation Protection Components .................................................. 3 .6.3-1
3.6.3.2 Safe Operating Ranges .................................................................... 3.6.3-2 •
3.6.4 Thermal Evaluation for Normal Conditions ofTranspo11- STC-LACBWR .. 3.6.4-1
3 .6.4.1 Thermal Models .............................................................................. 3 .6.4-1
3 .6.4.2 Maximum Temperatures ................................................................. 3 .6.4-7
3.6.4.3 Minimum Temperatures .................................................................. 3.6.4-8
3.6.4.4 Maximum Internal Pressure ............................................................ 3.6.4-8
3.6.4.5 Maximum Thermal Stresses ......................................................... 3.6.4-13
3.6.4.6 Summary of STC-LACBWR Performance for Normal Conditions
of Transport. .................................................................................. 3 .6 .4-13
3.6.5 Hypothetical Accident Thermal Evaluation - STC-LACBWR ................... 3.6.5-1
• 3-ii
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NAC-STC SAR
Docket No. 71-9235
Table of Contents
(Continued)
March 2017
Revision 18
3.7 Thermal Evaluation- STC-WVDP .............................................................................. 3.7-1
3.7.1 Discussion - STC-WVDP HL W Overpack and Contents ................................ 3.7.1-1
3.7.2 Summary ofThermal Properties ofMaterials-STC- WVDP ......................... 3.7.2-1
3.7.3 Technical Specifications for Components- STC - WVDP ........................... 3.7.3-1
3.7.3.1 Radiation Protection Components .................................................. 3.7.3-1
3.7.3.2 Safe Operating Ranges .................................................................... 3.7.3-1
3.7.4 Thermal Evaluation for Normal Conditions ofTranspo11-MPC - WVDP .... 3.7.4-1
3.7.4.1 Thermal Model.. .............................................................................. 3.7.4-1
3.7.4.2 Maximum Temperatures ................................................................. 3.7.4-3
3.7.4.3 Minimum Temperatures .................................................................. 3.7.4-4
3. 7.4.4 Maximum Internal Pressure ............................................................ 3. 7.4-4
3.7.4.5 Maximum Thermal Stresses ........................................................... 3.7.4-4
3.7.4.6 Summary of STC- WVDP Performance for Normal Conditions
of Transport ..................................................................................... 3. 7 .4-4
3. 7 .5 Hypothetical Accident Thermal Eva! uation - STC- WVDP ......................... 3. 7 .5-1
3.8 Thermal Evaluation - STC-High Burnup Directly Loaded Fuel (STC-HBU) ............. 3.8-1
3.8.1 Discussion - STC-HBU and Contents ........................................................... 3.8.1-1
3.8.2 Summary of Thermal Properties of Materials - STC-HBU .......................... 3.8.2-1
3.8.3 Technical Specifications for Components - the STC-HBU .......................... 3.8.3-1
3.8.3.1 Radiation Protection Components .................................................. 3.8.3-1
3.8.3.2 Safe Operating Ranges .................................................................... 3.8.3-2
3.8.4 Thermal Evaluation for Normal Conditions of Transport - the STC-HBU .. 3.8.4-1
3.8.4.1 Thermal Models .............................................................................. 3.8.4-l
3.8.4.2 Maximum Temperatures ................................................................. 3.8.4-6
3.8.4.3 Minimum Temperatures .................................................................. 3.8.4-8
3.8.4.4 Maximum Thermal Stresses ........................................................... 3.8.4-8
3.8.4.5 Summary of the STC-HBU Performance for Normal Conditions of
Transpo1i ......................................................................................... 3.8.4-8
3.8.5 Hypothetical Accident Thermal Evaluation - the STC-HBU ........................ 3.8.5-1
3.8.6 Maximum Pressure During Normal and Hypothetical Accident Conditions (HAC)
of Transport .................................................................................................... 3.8.6-1
3-iii
NAC-STC SAR
Docket No. 71-9235, Revision 17B
List of Figures
Figure 3.1-1 Definition of the Gap between the Basket and the Inner Shell
June2018
Revision 18B
for the Horizontal Position of the Cask ........................................................... 3 .1-10
Figure 3.1-2 Definition ofthe Gap between the Yankee-MPC Basket, Canister,
and the Inner Shell for the Horizontal Position of the NAC-STC .................. 3 .1-10
Figure 3.1.3 Basket Orientation and Gap between the CY-MPC Basket, Canister
and the Inner Shell for the Horizontal Position of the NAC-STC .................. 3 .1-11
Figure 3.2-1 Radial Temperature Profile versus NS-4-FR Thermal Conductivity
for Directly Loaded Fuel.. ................................................................................. 3.2-9
Figure 3.3-1 NS-4-FR Developer's Test Results Letter. ........................................................ 3.3-5
Figure 3.3-2 JAPC NS-4-FR Technical Data ........................................................................ 3.3-6
Figure 3.4-1 Three-Dimensional ANSYS Model for Directly Loaded Fuel ....................... 3.4-46
Figure 3.4-2 Design Basis Directly Loaded PWR Fuel Assembly
Axial Flux Distribution ................................................................................... 3.4-47
Figure 3.4-3 Horizontal View of the ANSYS Model for Directly Loaded Fuel
Containing the Support Disk, Fuel Assembly Elements and Shell. ................ 3.4-48
•
Figure 3.4-4 Detailed View of a Portion of the ANSYS Directly Loaded •
Fuel Basket Model .......................................................................................... 3 .4-49
Figure 3.4-5 Isometric View of the Directly Loaded Fuel Elements
for the Thermal Model .................................................................................... 3.4-50
Figure 3.4-6 Isometric View of the 180-Degree Section Cask Thermal
Model for Directly Loaded Fuel ..................................................................... 3.4-51
Figure 3.4-7 Detailed View of Basket and Shells of the 180-Degree
Section Cask Thermal Model for Directly Loaded Fuel.. ............................... 3.4-52
Figure 3.4-8 Plan View of the Directly Loaded Fuel 180-Degree Section
Cask Thern1al Model ....................................................................................... 3.4-53
Figure 3.4-9 Directly Loaded Fuel Assembly Thermal Model ........................................... 3.4-54
Figure 3 .4-10 Detailed View of a Single Fuel Rod in the Directly Loaded Fuel
Assembly Thermal Model ............................................................................... 3.4-55
Figure 3.4-11 Directly Loaded Fuel Basket Temperature Distribution for the Steel
Support Disk with Helium .............................................................................. 3.4-56
Figure 3.4-12 Directly Loaded Fuel Basket Temperature Distribution for the
Aluminum Heat Transfer Disk with Helium .................................................. 3.4-57
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NAC-STC SAR June 2018
Revision l 8B Docket No. 71-9235, Revision l 7B
Figure 3.4-13
Figure 3.4-14
Figure 3.4-15
Figure 3 .4-16
Figure 3.4-17
Figure 3 .4-18
Figure 3 .4-19
Figure 3.4-20
Figure 3.4-21
Figure 3.4-22
Figure 3.4-23
Figure 3.4-24
Figure 3 .4-25
Figure 3.4-26
Figure 3.4-27
Figure 3.4-28
Figure 3.4-29
Figure 3.4-30
Figure 3.4-31
Figure 3.4-32
Figure 3.4-33
Figure 3.4-34
List of Figures
(Continued)
Directly Loaded Fuel Basket Temperature Distribution for the Steel
Support Disk with Air. .................................................................................... 3.4-58
Directly Loaded Fuel Basket Temperature Distribution for the
Aluminum Heat Transfer Disk with Air ......................................................... 3.4-59
Isometric View of Quarter Symmetry Heat-up Transient
Model for Directly Loaded Fuel ..................................................................... 3 .4-60
Heat-up Transient Thermal Response of the Directly Loaded
Basket Aluminum Disk ................................................................................... 3.4-61
Heat-up Transient Average Temperature Response for Directly
Loaded Fuel Basket Aluminum Disk and Inner Shell Wall ........................... 3.4-62
Heat-up Transient Thermal Response for the Directly
Loaded Fuel Basket Steel Support Disk ......................................................... 3 .4-63
Heat-up Transient Average Temperature Response for the Directly
Loaded Fuel Basket Steel Support Disk and Inner Shell Wall ....................... 3.4-64
Three-Dimensional ANSYS Model for Yankee-MPC Canistered Fuel ......... 3.4-65
Design Basis Yankee Class Canistered Fuel Assembly Axial
Power Distribution .......................................................................................... 3.4-66
Fuel Assembly Model for Yankee-MPC Canistered Fuel .............................. 3.4-67
Fuel Tube Model for Yankee-MPC Canistered Fuel ...................................... 3.4-68
Two-Dimensional Yankee Reconfigured Fuel Assembly Model ................... 3.4-69
Yankee Damaged Fuel Locations in the Three-Dimensional Cask Model... .. 3.4-70
Three-Dimensional Cask Thermal Model for the CY-MPC ........................... 3.4-71
Three-Dimensional Cask Thermal Model for CY-MPC - Cross-Section ...... 3.4-72
Design Basis Connecticut Yankee Fuel Assembly Axial
Power Distribution .......................................................................................... 3 .4-73
Quarter-Symmetry Connecticut Yankee Fuel Assembly Model .................... 3.4-74
Fuel Tube Model for Connecticut Yankee Canistered Fuel ........................... 3.4-75
CY-MPC GTCC Transport Configuration Finite Element Model ................. 3 .4-76
CY-MPC GTCC Thermal Model Cross-Section ............................................ 3.4-77
Personnel Barrier Thermal Model .................................................................. 3 .4-78
Temperature Results for the Personnel Barrier. .............................................. 3 .4-79
3-v
NAC-STC SAR
Docket No. 71-9235
List of Figures
(Continued)
Figure 3.5-1 NAC-STC Hypothetical Accident Conditions ANSYS Model
March 2017
Revision 18
for Directly Loaded Fuel.. ............................................................................... 3 .5-11
Figure 3.5-2 NAC-STC Hypothetical Accident Conditions Temperature History
for the Directly Loaded Basket ....................................................................... 3.5-12
Figure 3.5-3 NAC-STC Hypothetical Accident Conditions Temperature History
for CY-MPC Fuel ........................................................................................... 3.5-15
Figure 3.6-1 Three-Dimensional ANSYS Model for STC-LACBWR ........................... 3.6.4-14
Figure 3.6-2 Three-Dimensional Cask Thermal Model for STC-LACBWR-
Cross-Section ............................................................................................... 3.6.4-15
Figure 3.6-3 Damaged Fuel Locations in the Three-Dimensional Cask and
Canister Model (Cross-Section) ................................................................... 3.6.4-16
Figure 3.6-4 Design Basis LACBWR Fuel Assembly Axial Power Distribution ............ 3.6.4-17
Figure 3.6-5 Fuel Assembly Model for LACBWR Fuel .................................................. 3.6.4-18
Figure 3.6-6 Two-Dimensional MPC-LACBWR Fuel Tube Model
(Standard Fuel Tube with BORAL Plate) .................................................... 3.6.4-19
Figure 3.6-7 Two-Dimensional MPC-LACBWR Fuel Tube Model
(Standard Fuel Tube without BORAL Plate) ............................................... 3.6.4-20
Figure 3.6-8 Two-Dimensional MPC-LACBWR Fuel Tube Model
(Fuel Tube in the Slots Containing DFC with BORAL) ............................. 3.6.4-21
Figure 3.6-9 Two-Dimensional MPC-LACBWR Fuel Tube Model
(Fuel Tube in the Slots Containing DFC without BORAL) ........................ 3.6.4-22
Figure 3.7-1 Three-Dimensional ANSYS Model for STC-WVDP ................................... 3.7.4-6
Figure 3.7-2 Three-Dimensional Model for STC- WVDP- Cross-Section ...................... 3.7.4-7
Figure 3.8-1 Configuration Definition for the STC-HBU ................................................ 3.8.4-10
Figure 3.8-2 Three-Dimensional ANSYS Model for the STC-HBU ............................... 3.8.4-11
Figure 3.8-3 Three-Dimensional Cask Thermal Model for the STC-HBU - Cross-Section
(Configuration 8) ......................................................................................... 3.8.4-12
Figure 3.8-4 HBU Fuel Assembly Axial Power Distribution ........................................... 3.8.4-13
Figure 3.8-5 Fuel Assembly Model for the STC-HBU Fuel ............................................ 3 .8.4-14
Figure 3.8-6 Two-Dimensional Fuel Tube Model for the STC-HBU .............................. 3.8.4-15
3-vi
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NAC-STC SAR June 2018
Revision 18B Docket No. 71-9235, Revision 17B
List of Tables
Table 3.1-1 Thermal Analysis Bounding Conditions - Normal Transport
Conditions ....................................................................................................... 3 .1-12
Table 3.2-1 Thermal Properties of Solid Neutron Shield (NS-4-FR) ................................ 3.2-10
Table 3.2-2 Thermal Prope1iies of Stainless Steel ............................................................. 3 .2-10
Table 3.2-3 Thermal Prope1iies of Chemical Copper Lead ............................................... 3 .2-11
Table 3.2-4 Thermal Properties of Type 606 l-T6 and 606 l-T65 I Aluminum
Alloy ............................................................................................................... 3.2-11
Table 3.2-5 Thermal Prope1iies of Helium ........................................................................ 3.2-12
Table 3.2-6 Thermal Prope1iies of Dry Air. ....................................................................... 3 .2-12
Table 3.2-7 Thermal Properties of Copper ......................................................................... 3 .2-12
Table 3.2-8 Thermal Properties of B4C .............................................................................. 3 .2-13
Table 3.2-9 Thermal Properties of Zircaloy and Zircaloy-4 Cladding ............................... 3 .2-13
Table 3.2-10 Thermal Properties of Fuel (U02) ................................................................... 3 .2-13
Table 3.2-11 Thermal Properties ofBORAL and Talbor Sheet... ........................................ 3.2-14
Table 3.4-1 Maximum Component Temperatures - Normal Transport
Conditions, Maximum Decay Heat and Maximum Ambient
Temperature - Directly Loaded and Canistered Configurations ..................... 3 .4-80
Table 3.4-2 Maximum Component Temperatures - Normal Transport
Conditions, Maximum Decay Heat, Minimum Ambient
Temperature - Directly Loaded and Canistered Configurations ..................... 3.4-81
Table 3.4-3 Maximum Component Temperatures - Normal Transport Conditions,
Maximum Decay Heat, Low Ambient, for Directly Loaded Fuel .................. 3.4-82
Table 3.4-4 NAC-STC Thermal Performance Summary for Normal
Conditions of Transport .................................................................................. 3 .4-83
Table 3.4-5 Maximum Cask Component Temperatures in Normal Conditions
of Transport ..................................................................................................... 3 .4-84
Table 3.4-6 Maximum Component Temperatures for Yankee-MPC Damaged Fuel ........ 3.4-85
Table 3.4-7 Maximum Component Temperatures for CY-MPC Damaged Fuel.. ............. 3.4-85
Table 3.4-8 Westinghouse 15 X 15 Fuel Assembly Characteristics .................................. 3.4-85
Table 3.4-9 Directly Loaded Fuel Basket Component Volumes ....................................... 3.4-86
Table 3.5-1 Maximum Component Temperatures - Hypothetical Accident
Conditions Fire Transient ............................................................................... 3 .5-16
Table 3.6-1 Thermal Analysis Bounding Conditions - Normal Transport
Conditions ...................................................................................................... 3 .6.1-4
3-vii
NAC-STC SAR
Docket No. 71-9235
List of Tables (continued)
March 2017
Revision 18
Table 3.6-2 Thermal Properties of Helium ....................................................................... 3.6.2-2
Table 3.6-3 Gaps within the STC-LACBWR Three-Dimensional Thermal Model ......... 3.6.2-3
Table 3.6-4 Maximum Component Temperatures - Normal Transport Conditions,
Maximum Decay Heat and Maximum Ambient Temperature -
STC-LACBWR ............................................................................................ 3.6.4-23
Table 3.6-5 Maximum Component Temperatures - Normal Transport Conditions,
Maximum Decay Heat, Minimum Ambient Temperature -
STC-LACBWR ............................................................................................ 3.6.4-24
Table 3.6-6 Maximum Temperature of the Fuel, Basket, and Canister- Hypothetical
Table 3.7-1
Table 3.7-2
Table 3.7-3
Table 3.7-4
Accident Condition Fire Transient.. ............................................................... 3.6.5-3
Thermal Analysis Bounding Conditions - Normal Transport
Conditions ...................................................................................................... 3.7.1-3
Thermal Properties of Glass ........................................................................... 3. 7 .2-2
Gaps within the STC-WVDP Three-Dimensional Thermal Model.. ............. 3.7.2-2
Maximum Component Temperatures - Normal Transport Conditions,
•
Maximum Decay Heat and Maximum Ambient Temperature - •
STC-WVDP ................................................................................................... 3.7.4-8
Table 3.7-5 Maximum Component Temperatures - Normal Transport Conditions,
Maximum Decay Heat, Minimum Ambient Temperature -
STC- WVDP .................................................................................................. 3.7.4-9
Table 3.7-6 Maximum Temperature of the HLW and Contents, Basket, and HLW
Overpack - Hypothetical Accident Condition Fire Transient ......................... 3.7.5-2
Table 3.8-1 Thermal Analysis Bounding Conditions - Normal Transpo11 Conditions .... 3.8.1-4
Table 3.8-2 Minimum Thermal Conductivities of Neutron Absorber (MMC, 45wt%
B4C) ............................................................................................................... 3.8.2-2
Table 3.8-3 Gaps within the STC-HBU Three-Dimensional Thermal Model .................. 3.8.2-3
Table 3.8-4 Maximum Component Temperatures-Normal Transport Conditions,
Maximum Decay Heat, Maximum Ambient Temperature, among Three
Configurations - the STC-HBU ................................................................... 3 .8.4-16
Table 3.8-5 Maximum Component Temperatures - Normal Transport Conditions,
Maximum Decay Heat, Minimum Ambient Temperature, among Three
Configurations - the STC-HBU ................................................................... 3.8.4-17
Table 3.8-6 Maximum Temperature of the STC-HBU - Hypothetical Fire Accident
Condition ........................................................................................................ 3 .8.5-2 •
3-viii
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NAC-STC SAR
Docket No. 71-9235, Revision 17B
June 2018
Revision 18B
analysis of the cask heat-up condition has been performed using the ANSYS finite element
program. The model, Figure 3.4-15, represents a quarter symmetry slice from the center section
of the cask and includes the fuel assemblies, fuel tubes, steel support disks, aluminum heat
transfer disks, and cask body wall. Each of these areas and components are modeled using the
material properties and detail represented in heat transfer finite element models discussed earlier,
with added heat transfer enhancement representing radiation between the tubes, and between the
tubes and the cask inner shell in the spaces between the support disks and the aluminum heat
transfer disks. In order to capture the influence of the initial vacuum drying process, the boundary
conditions starting the transient represented all components at 70°F, fuel assembly design basis
heat load of 0.85 kilowatts, and the cavity evacuated. At twenty-four hours into the transient,
helium was added to the model representing the normal operating procedure of back filling the
cavity with helium following completion of the drying process.
Component temperature profiles were obtained for each time step through cask steady state
conditions. Figures 3.4-16, 3.4-17, 3.4-18, and 3.4-19 present the transient temperature results
for the aluminum heat transfer disk; aluminum heat transfer disk average temperature and average
inner shell temperature; support disk; and support disk average disk temperature and average
inner shell temperature, respectively.
It is concluded from these results that a steady state heat flow is established throughout the cask
at approximately I 00 hours after fuel load with actual peak temperatures reached at about 240
hours after fuel load. Temperatures from this analysis are used as input loading to evaluate the
potential for basket and cask wall interference resulting from thermal expansion.
For canistered fuel, the canister configuration has been evaluated to ensure that the canister at the
steady state hot condition can be installed in an NAC-STC at the steady state cold condition. The
cold condition temperature is limited to 0°F, since this is the limiting temperature for operation
of the transfer cask. The transfer cask is used to install the canister in the NAC-STC .
3.4-45
NAC-STC SAR
Docket No. 71-9235, Revision l 7B
Figure 3.4-1 Three-Dimensional ANSYS Model for Directly Loaded Fuel
3.4-46
June 2018
Revision l 8B •
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•
NAC-STC SAR
Docket No. 71-9235, Revision l 7B
June 2018
Revision l 8B
Figure 3.4-2 Design Basis Directly Loaded PWR Fuel Assembly Axial Flux Distribution
z 0
8 IX ...)
w ::> L&.
~ .::. u ,c(
u.. 0 I-X 0 w ::x::
,oo,;
75~
50%
25~
o"
(100~ •. 53)
•·
(0%..53)
0 . 1 .2 .3 .4 .5 .6 . 7 .8 .9 1.0 1.1
RELA 11VE ?OWER
3.4-47
(80%.1.1)
(15%, 1.1)
NAC-STC SAR
Docket No. 71-9235, Revision l7B June 2018
Revision l 8B
Figure 3.4-3 Horizontal View of the ANSYS Model for Directly Loaded Fuel Containing the
Support Disk, Fuel Assembly Elements and Shell
3.4-48
Fuel
Basket
Basket/Inner Shell Gap
Cask Inner Shell
Lead Ga11111a Shield
Cask Outer Shell
Radial Neutron Shield
Neutron Shield Shell
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•
NAC-STC SAR
Docket No. 71-9235, Revision 17B
June2018
Revision 18B
Figure 3 .4-4 Detailed View of a Portion of the AN SYS Directly Loaded Fuel Basket Model
Fuel
Composite Fuel Tube (stainless steel, BORAL
Gap between Fuel & Fuel Tube & Air Gaps)
3.4-49
ii
Basket
NAC-STC SAR
Docket No. 71-9235, Revision l7B
June2018
Revision l 8B
Figure 3.4-5 Isometric View of the Directly Loaded Fuel Elements for the Thermal Model
3.4-50
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NAC-STC SAR
Docket No. 71-9235, Revision l7B
June2018
Revision l 8B
Figure 3.4-6 Isometric View of the 180-Degree Section Cask Thermal Model for Directly
Loaded Fuel
3.4-51
NAC-STC SAR
Docket No. 71-9235, Revision 17B
June2018
Revision 18B
Figure 3.4-7 Detailed View of Basket and Shells of the 180-Degree Section Cask Thermal
Model for Directly Loaded Fuel
Stainless Steel Support Disk
Aluminum Heat Transfer Fin
Basket/Inner Shell Gap
Inner Shell
Ganna Shield
Gap
Outer Shell
Radial Neutron Shield
Radial Neutron Shield Shell
3.4-52
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NAC-STC SAR
Docket No. 71-9235, Revision 17B
June 2018
Revision 18B
Figure 3.4-8 Plan View of the Directly Loaded Fuel 180-Degree Section Cask Thermal Model
-------Tl--- Maximum Gap
3.4-53
( 0 .13 inches)
Basket
Basket Inner Shell Contact Surface
Inner Shell
NAC-STC SAR
Docket No. 71-9235, Revision 17B
Figure 3.4-9 Directly Loaded Fuel Assembly Thermal Model
Plane of S}'ITl11etry
Exterior of Fuel Assembly Model
3.4-54
June 2018
Revision 18B
Center of Fuel Assembly
Plane of Syn111etry
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NAC-STC SAR
Docket No. 71-9235, Revision 17B
June 2018
Revision 18B
Figure 3.4-10 Detailed View of a Single Fuel Rod in the Directly Loaded Fuel Assembly
Thermal Model
Fuel Cladding Cavity Gas
3.4-55
NAC-STC SAR
Docket No. 71-9235, Revision 17B
June2018
Revision 18B
Figure 3.4-11 Directly Loaded Fuel Basket Temperature Distribution for the Steel Support
Disk with Helium
SteeJ 01sk
Transport/Hel1um
Maximum Gap ~-_,...20
7 93
.:10 EJD68 38 EJ EJ '----79~ 95
EJEJEJ6~ 23
/ "
79 87
Region of Contact'-- 01
~ (Temperatures in °F)
3.4-56
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•
NAC-STC SAR
Docket No. 71-9235, Revision 17B
June 2018
Revision 18B
Figure 3.4-12 Directly Loaded Fuel Basket Temperature Distribution for the Aluminum Heat
Transfer Disk with Helium
35 39
Aluminum Fin Region
Transport/Helium
(Temperatures in °F)
Maximum Gap __ ___,....16
B 85
04 46 60
29 72 87
01 43 57
EJ .___7 2___...__.__..
80
11
3.4-57
NAC-STC SAR
Docket No. 71-9235, Revision l 7B
June 2018
Revision l 8B
Figure 3.4-13 Directly Loaded Fuel Basket Temperature Distribution for the Steel Support
Disk with Air
Maximum Gap ____ ,28
93
10 55 70
51 52 37 81 97
52 67
80 89
Steel Disk Transport/AIR
Region of Contact
~ 03
(Temperatures in °F)
3.4-58
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NAC-STC SAR
Docket No. 71-9235, Revision 17B
June 2018
Revision 18B
Figure 3.4-14 Directly Loaded Fuel Basket Temperature Distribution for the Aluminum
Heat Transfer Disk with Air
Aluminum Fin Transport/AIR
Maximum Gap
06
32
Reg1on of Contact
(Temperatures in °F)
3.4-59
__ _.,..16
86
63
75 90
60
74 82
NAC-STC SAR
Docket No. 71-9235, Revision l7B
June2018
Revision l 8B
Figure 3.4-15 Isometric View of the Quaiter Symmetry Heat-up Transient Model for
Directly Loaded Fuel
3.4-60
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•
NAC-STC SAR
Docket No. 71-9235, Revision 17B
June2018
Revision I SB
Figure 3.4-16 Heat-up Transient Thermal Response of the Directly Loaded Basket
Aluminum Disk
TEMP 800
720
640
560
.:180
.:100
320
2.ao
160
80
~~
~ -;:::::------
If ~ ~ ~ ~ l ---
~ 'l ~ ~
r
' fl 0
0 80 160 40 120
3.4-61
240 200 280
.1 ~ ~L Q ~ - -
158 AL .1.1
1 F;] f. I 15 ~ _::;i AL .1 ::l
:, 34 AL 23
::, li3~~ §i
320 360
HRS
400
NAC-STC SAR
Docket No. 71-9235, Revision 17B
June2018
Revision 18B
Figure 3.4-17 Heat-up Transient Average Temperature Response for Directly Loaded Fuel
Basket Aluminum Disk and Inner Shell Wall
TEMP 500
A50 ~
Ll DO /
350 I
I 300
250
200
150
100
I ~ v-
I V
/ I I V
II V
50
0 0 80 150
40 120
3.4-62
240 200 280
l'.'."[NS-T AV
TSHL-T AV
320 360
4
HRS
00
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•
NAC-STC SAR
Docket No. 71-9235, Revision 17B
June 2018
Revision 18B
Figure 3.4-18 Heat-up Transient Thermal Response for the Directly Loaded Fuel Basket
Steel Suppo1i Disk
TEMP 800
720
640
1 ~ ~~ 0 560
.::180
400
320
2~0
160
80
r::::::-- 170 s~ 1 1
~ ~ 169 ss 15
IP, ~ [:::::::= u--::r --;:J..; r-::J
') g5 55 23 ---~ / J ? 95 s~ 28 - 1q335~ 32
~ v/% ~ ...... -::i 37 S= .35 -~
y ~
' ~ y 0 HRS
0 80 160 240 320 400 40 120 200 280 360
3.4-63
NAC-STC SAR June 2018
Revision 18B Docket No. 71-9235, Revision 17B
Figure 3.4-19 Heat-up Transient Average Temperature Response for the Directly Loaded
Fuel Basket Steel Supp011 Disk and Inner Shell Wall
TEMP 500
n1SK-T AV Ll50 ----LIDO
/ /
350
300
250
200
150
100
I I
I T SHL-T AV
~ ~
I /
/ I V
I
II V
50
0 HRS 0 80 160 240 320 ADO
-40 120 200 280 360
3.4-64
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NAC-STC SAR
Docket No. 71-9235, Revision l7B
June2018
Revision l 8B
Figure 3.4-20 Three-Dimensional ANSYS Model for Yankee-MPC Canistered Fuel
Spacer
Canister Bottom Plate
3.4-65
Gap
Neutron Shield Shell
Spacer
Outer Shell
Lead
NAC-STC SAR
Docket No. 71-9235, Revision 17B June 2018
Revision 18B
Figure 3.4-21 Design Basis Yankee Class Canistered Fuel Assembly Axial Power Distribution
L
0
+:>
l .
U 1.0 d
LL
C
0
.9
.8
2
8
6
4
I
I \ r1 \
7 \ .,....( .7
L
~
(/")
z7 \ 0
'a.,
3: 0
0...
a; ::,
LL
.4
.J
.2
. l
6 /
8
6
4
2
0 0 .2 .4 .6 8
. l .J .5 .7 .9
Fraction of Active Fuel Length
3.4-66
\ \
1
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•
NAC-STC SAR
Docket No. 71-9235, Revision l 7B
Figure 3.4-22 Fuel Assembly Model for Yankee-MPC Canistered Fuel
3.4-67
June 2018
Revision 18B
NAC-STC SAR
Docket No. 71-9235, Revision 17B
June20]8
Revision l SB
Figure 3.4-23 Fuel Tube Model for Yankee-MPC Canistered Fuel
Fuel Tube
Helium Gap
Helium Gap
Stainless Steel Clad for BORAL Plate
Helium Gap
Aluminum Clad ofBORAL Plate
Core Matrix ofBORAL Plate
Aluminum Clad ofBORAL Plate
3.4-68
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•
NAC-STC SAR
• Docket No. 71-9235, Revision l 7B
Figure 3.4-24 Two-Dimensional Yankee Reconfigured Fuel Assembly Model
Shell Casing
1Desi gn Case :rfa_tr
•
Fuel Rod Helium Tube
• 3.4-69
June 2018
Revision I SB
ANSYS APR 9 16:34:1 PLOT NO. ELEMENT TYPE
zv D1ST=2.1 XF YF
Shell Casing
Radiation Element (Typ.)
NAC-STC SAR
Docket No. 71-9235, Revision l7B June 2018
Revision l 8B
Figure 3.4-25 Yankee Damaged Fuel Locations in the Three-Dimensional Cask Model
Damaged fuel can is restricted to these positions
Debris of the 20 failed fuel rods is concentrated in the center 7 .1 inches of the active fuel region
3.4-70
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NAC-STC SA R
Docket o. 71-9235 , Rev ision 17B
June2018
Rev ision I 8B
Figure 3.4-26 Three- Dimensional Cask Thermal Model fo r the CY-MPC
Cask Outer Bottom
r-Canister Bottom Plate I I
Fuel Basket Bottom Weldment
Lower Neut ran Shield
Support Disks
Heat Transfer
Neutron Shield Shell
Radial Neutron Shield
Outer Shell
Gamma Shield
Inner Shell
Canister Shell
Fuel Basket Top Weldment
Note : canister fill gas , cask fill gas, and bottom spacer not shown for clariq.1 .
Cask Inner Lid
Upper Neutron Shield
Cask Outer Lid
3.4-7 1
Shield Lid
Lid
NAC-STC SAR
Docket No. 71-9235 , Revisi on 178
June20 18
Revi s ion I 8B
Figure 3.4-27 Three-Dimensiona l Cask Thermal Model for CY-MPC - Cross-Section
Oversize FUel Assembly
~ Neutron Shield Shell
Neutron Shield
Gamma Shield
Cask Inner Shell Fuel Tubes
Helium (Cask Gas)
Canister Shell
No Gaps
Support Disk/Heat Transfer Disk/Helium
3.4-72
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•
NAC-STC SAR
Docket No. 71-9235 , Rev ision l 7B
Ju ne 20 18
Rev ision I 8B
Figure 3.4-28 Des ign Basis Connecticut Yankee Fuel Assembl y Axial Power Distribution
1.2
1 . 1 :1. 1
I L I I I
0. , - -----. ---.-::, C: L. ::, al I
0.6 T
[_
Q)
> ... ra I
Q) 0.4
0 .4:33 a::: -,--
0.2
0
0 0.1 0.2 0 .3 0 .4 0.5 0.6 0 .7 0.8 0.9
Fraction of Core Height
3.4-73
NAC-STC SAR
Docket No. 71-9235 , Rev ision 17B
F igure 3.4-29 Quarter-Symmetry Connect icut Yankee Fue l Assembly Model
Quarter - Symmetry Bounda r y
Center of Fuel Assembly
Quarter-Symmetry Boundary
3.4-74
Inner Surface of Fuel Tube
Fuel Cladding
Helium
June 20 18
Revi sion l 8B •
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NAC-STC SAR
Docket No. 71-9235 , Revision I 7B
June 2018
Revision I 8B
Figure 3.4-30
Heat Flu:~
Fuel Tube Model for Connecticut Yankee Cani stered Fuel
y
Core Matri:-: of BORAL Plate
Fuel Tube
Gap r sca,o,ess
• • ~
Steel Clad for BORAL Plate
I •
Helium Gap
Inner Edge of Support/Heat
/ Transfer Disk • Slot
L,__ _ __..._,__•_,___ __ -r-'-•---- -----x
Aluminum Clad of BORAL Plate
Radiation Link Element 1
3.4-75
NAC-STC SAR
Docket No. 71-9235 , Rev ision I 7B
June 20 18
Revision I 8B
Figure 3.4-31 CY-MPC GTCC Transport Configuration Finite Element Model
Neut ron Shield Shell
Neutron Shield
Cask Outer Shell
0.01 5 Air GaR
GTCC Canister
Tube Array Weldment
Shield Shell Weldment
GTCC Waste
3.4-76
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NAC-STC SAR
Docket No. 71-9235 , Revi sion l 7B
Figure 3.4-32 CY-MPC GTCC Therma l Model Cross-Section
Ne ut ron Shield Shell
Neut ron Shield
Cask Outer Shell
Lead
Cask Inner
Helium Fi ll Gas
-
-
Shield Shell Weldme 111
Tube Array We ld,
GTCC Waste Helium
----11 en t
and --Gas
-Tube Array We ld mcnt
GTCC Waste Helium
and --Gas
Tube Array Weldm ent
GTCC Waste Helium
Tube Array Weld
and --Gas
ment
~ ·--- -----
-
-
-
-
-i--
--
3.4-77
/
.--
--
June 20 18
Revi sion 18B
0.015AirGap
GTCC Canister Shell
GTCC Shield Shell Support Disk
Y (Radia l)
z J
Tran sport Cask and Basket Center Line
NAC-STC SAR
Docket No. 71-9235, Revision I 7B
F igure 3.4-33 Personnel Barrier Therma l Model
l 12.8 in .
!
Cask Surface
Location of
the Personnel
Barrier
3.4-78
11111111
lllli ll\11
~
~
lli, '\
I 'J
~,___r-
"r-"1--1--
r-r-r-
' '
I'-" i'-1'-
I'-..._
I'-
1' i'-1'-
I'-
June 20 18
Revi sion 18B
', i'-,1'-
1--'I'-
1--, I'-"
', 'i'-1--
" ' 'I'-
' 1'-, 'I'- ' 1,
1' " 1' ' 1'
' ' 1'
' '\
' ' ' 1'
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•
NAC-STC SAR
Docket No. 71-9235, Revision 17B
June 2018
Revision 18B
Figure 3.4-34 Temperature Results for the Personnel Barrier
Pl P2
Cl
Cask Surface
C2 _/
C3
Boundary Conditions (°F)
Location Cl C2 C3 Pl
Temperature 230 244 258 140
3.4-79
P3
P4
PS
Personnel Barrier
Calculated Temperature (°F)
P2 P3 P4
105 126 127
PS
JOO
NAC-STC SAR
Docket No. 71-9235, Revision l 7B
June 2018
Revision l 8B
Table 3.4-1 Maximum Component Temperatures-Normal Transport Conditions, Maximum
Decay Heat and Maximum Ambient Temperature-Directly Loaded and
Canistered Configurations
Conditions: 100°F Ambient Temperature, Full Insolation, Decay Heat Load: 22.1 kW for Uncanistered Fuel; 12.5
kW for Yankee-MPC Canistered Fuel; 17 kW for Connecticut Yankee-MPC Canistered Fuel
Canistered Fuel
Directly Loaded Yankee-MPC CY-MPC
Cavity Gas Cavity Gas Cavity Gas
Component Air (°F) Helium (°F) Notes Helium (°F) Notes Helium (°F) Notes
Outer Lid 0-ring 178 176 (1) 176 (4) 157 (6)
Port Cover 0-rings 211 210 (1) 210 (4) 179 (7)
Inner Lid and Port Cover
Plate 0-rings 190 189 (1) 189 (4) 179 (7)
Cask Radial Outer Surface 241 243 (2) 243 (5) 258 (8)
Top Neutron Shield 181 175 (1) 175 (4) 168 (8)
Radial Neutron Shield 284 285 (2) 270 (5) 288 (8)
Lead Gamma Shield 314 315 (2) 281 (5) 300 (8)
Aluminum Disk Exterior 338 337 (2) --- --- 331 (8)
Aluminum Disk Interior 491 487 (2) 536 (5) 534 (8)
Steel Support Disk Exterior 356 344 (2) --- --- 324 (8)
Steel Support Disk Interior 498 495 (2) 539 (5) 536 (8)
Canister Shell --- --- --- 338 (5) 351 (8)
Canister Lid --- --- --- 209 (5) 220 (8)
Canister Bottom Plate --- --- --- 255 (5) 347 (8)
Maximum Fuel Rod Cladding 588 544 (3) 575 (5) 611 (8)
Notes: (1) Temperatures are determined from the analysis of the three-dimensional quarter symmetry model of the entire cask (directly loaded fuel).
(2)
(3)
(4)
(5) (6)
(7)
(8)
Temperatures are determined from the analysis of the three-dimensional 180-degree section model of the entire cask ( directly loaded fuel). Temperatures are determined from the analysis of the two-dimensional detailed model of the fuel assembly (directly loaded fuel). Component not explicitly modeled in the 3-D model for Yankee-MPC canistered fuel. Temperature results from the helium case of the directly loaded fuel used (conservative).
Temperatures are determined from the 3-D model for Yankee-MPC canistered fuel. Not explicitly modeled-maximum temperature of cask outer lid from 3-D model for CY-MPC presented. Not explicitly modeled-maximum temperature of the cask top forging/cask lids from 3-D model for CY-MPC presented. Temperatures are determined from the 3-D model for CY-MPC canistered fuel.
3.4-80
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NAC-STC SAR
Docket No. 71-9235, Revision 17B
June 2018
Revision 18B
Table 3.4-2 Maximum Component Temperatures - Normal Transport Conditions, Maximum
Decay Heat, Minimum Ambient Temperature - Directly Loaded and Canistered
Configurations
Conditions: -40°F Ambient Temperature, No Insolation, Decay Heat Load: 22.1 kW for Uncanistered Fuel; 12.5 kW
for Yankee-MPC Canistered Fuel; 17 kW for Connecticut Yankee-MPC Canistered Fuel
Directly Loaded Yankee-MPC CY-MPC
Components Air (°F) Notes Helium (°F) Notes Helium (°F) Notes
Outer Lid 0-ring 125 --- 125 (2) 86 (4)
Inner Lid and Inner Lid Port 125 --- 125 (2) 89 (5)
Cover Plate 0-rings
Port Cover 0-rings 129 --- 129 (2) 170 (5)
Cask Radial Outer Surface 144 --- 116 (3) 162 (6)
Top Neutron Shield 131 --- 131 (2) 87 (6)
Radial Neutron Shield 181 --- 142 (3) 162 (6)
Lead Gamma Shield 215 --- 154 (3) 175 (6)
Fuel Basket Exterior 256 (1) --- --- --- ---Maximum Basket Web 399 (1) 431 (3) 428 (6)
Canister Shell --- --- 215 (3) 232 (6)
Canister Lid --- --- 71 (3) 92 (6)
Canister Bottom Plate --- --- 121 (3) 229 (6)
Fuel Rod Cladding 488 (1) 473 (3) 512 (6)
Notes: (1) Basket and fuel rod cladding temperatures are defined by adding the gradient result between the lead gamma shield and point of interest obtained from the 3-D directly loaded fuel model with air in the cavity (Table 3 .4-1 ).
(2) Component not explicitly modeled in the 3-D model for Yankee-MPC canistered fuel. Temperature results from the air case of the directly loaded fuel used (conservative).
(3) Temperatures obtained from 3-D model for Yankee-MPC canistered fuel. (4) Not explicitly modeled-maximum temperature of cask outer lid from 3-D model for Connecticut
Yankee-MPC presented. (5) Not explicitly modeled-maximum temperature of cask top forging/cask lids from 3-D model for
Connecticut Yankee-MPC presented. (6) Temperatures obtained from 3-D model for Connecticut Yankee-MPC canistered fuel.
3 .4-81
NAC-STC SAR
Docket No. 71-9235, Revision l7B
June2018
Revision J SB
Table 3.4-3 Maximum Component Temperatures - Normal Transport Conditions,
Maximum Decay Heat, Low Ambient, for Directly Loaded Fuel
Conditions: -20°F Ambient Temperature, 22.l kW Decay Heat and No Insolation
Component Temperature (°F)
Outer Lid 0-ring 161
Inner Lid and Port Cover Plate 0-rings 165
Port Cover 0-rings 165
Cask Radial Outer Surface 173
Top Neutron Shield 168
Radial Neutron Shield 211
Lead Gamma Shield 245
Fuel Basket Exterior1 286
Maximum Basket Web 1 429
Maximum Fuel Rod Cladding1 518
1 Basket and fuel rod cladding temperatures are defined by adding the gradient
result between the lead gamma shield and point of interest obtained from the
3-D finite element analysis with air in the cavity (Table 3.4-1).
3.4-82
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NAC-STC SAR
Docket No. 71-9235, Revision 17B
June 2018
Revision 18B
Table 3.4-4 NAC-STC Thermal Performance Summary for Normal Conditions of Transport
Temperature Range
Directly Loaded Canistered Fuel
Component Fuel Yankee-MPC CY-MPC Allowable Temperature
< 380°C-Uncanistered
Fuel Cladding1 309°c 302°c 322°c < 340°C-Yankee-MPC
< 341 °C-CY -MPC
Metallic 0-rings -40 to 190°F -40 to 190°F -40 to 218°F -40 to 500°F
Viton 0-rings -40 to l 90°F -40 to 190°F -40 to 218°F -40 to 400°F
PTFE O-rings2 -40 to 190°F -40 to l 90°F -40 to 218°F -40 to 650°F
Radial NS-4-FR Neutron
Shield -40 to 285°F -40 to 270°F -40 to 288°F -40 to 300°F
Lead Gamma Shield -40 to 3 l 5°F -40 to 281 °F -40 to 300°F -40 to 600°F
Aluminum Heat Transfer Disk -40 tO 491 Of -40 to 536°F -40 to 534°F -40 to 600°F
I. Allowable temperatures for uncanistered ( directly loaded) fuel and for Yankee fuel in the
Yankee-MPC are based on PNL-4835. The allowable temperature for Connecticut Yankee
fuel is based on the methodology of PNL-6364, to consider preferentially loaded
configurations of the CY-MPC system.
2. The safe operating range extends to 735°F. (Certified Test Report D9-3362-I, Applied
Technical Services, Inc., February 8, 1989.) An allowable temperature of 650°F is
conservatively applied .
3.4-83
NAC-STC SAR June 2018
Revision 18B Docket No. 71-9235, Revision 17B
Table 3.4-5
NAC-STC
Components
Bottom Plate
Bottom Forging
Transition Shell
Inner Shell
Outer Shell
Top Forging
Inner Lid
Outer Lid
Inner Lid Bolt
Outer Lid Bolt
Notes: (I)
(2)
(3)
(4)
(5)
(6)
(7)
(8)
Maximum Cask Component Temperatures in Normal Conditions of Transport
Directly Loaded Fuel Canistered Fuel
Cavity Gas Yankee-MPC CY-MPC
Air (°F) Helium (°F) Notes Helium (°F) Notes Helium (°F) Notes
350 333 (I) 333 (3) 347 (5)
417 393 (I) 393 (3) 347 (6)
300 300 (I) 300 (3) 331 (7)
331 331 (2) 311 (4) 331 (5)
292 293 (2) 276 (4) 294 (5)
211 210 (I) 210 (3) 218 (5)
223 210 (I) 210 (3) 217 (5)
178 179 (I) 176 (3) 216 (5)
190 189 ( 1) 189 (3) 217 (8)
178 176 ( 1) 176 (3) 216 (8)
Temperatures are determined from the analysis of the three-dimensional quarter symmetry model of
the entire cask.
Temperatures are determined from the analysis of the three-dimensional 180-degree section model
of the entire cask.
Component not explicitly modeled in the 3-D model for Yankee-MPC canistered fuel.
Temperature results from the helium case of the directly loaded fuel used (conservatively).
Temperatures are determined from the 3-D model for Yankee-MPC canistered fuel.
Temperatures are determined from the 3-D model for CY-MPC canistered fuel.
Not explicitly modeled-taken as the maximum temperature of the bottom plate from the 3-D
model for CY-MPC canistered fuel.
Not explicitly modeled-taken as the maximum temperature of the inner shell from the 3-D model
for CY-MPC canistered fuel.
Not explicitly modeled-taken as the maximum temperature of the inner and outer lids from the
3-D model for CY-MPC canistered fuel.
3.4-84
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NAC-STC SAR
Docket No. 71-9235, Revision 17B
June 2018
Revision 18B
Table 3.4-6 Maximum Component Temperatures for Yankee-MPC Damaged Fuel
Contents Heat Transfer Disk Damaged Fuel Fuel Cladding
Four damaged fuel and 32 548 544 586
intact fuel assemblies (°F)
Allowable Temperatures 600 NIA 644
(Table 3 .4-4) (°F)
Table 3.4-7 Maximum Component Temperatures for CY-MPC Damaged Fuel
Maximum Temperature1 (°F)
Design Damaged Fuel Reconfigured Fuel Fuel Rod
Condition Can Assembly Cladding2
100°F Ambient 611 611 611
-40°F Ambient 512 512 512
Bounding temperatures are taken from the maximum fuel cladding temperature for the design basis fuel
(654 watts per assembly) .
2 The allowable fuel cladding temperature is 646°F (341 °C) (See Table 3 .4-4).
Table 3.4-8 Westinghouse 15 x 15 Fuel Assembly Characteristics
Parameter Units Value
Number of Fuel Rods -- 204
Fuel Rod Outer Diameter inch 0.422
Fuel Pellet Diameter inch 0.3659
Fuel Rod Clad Inner Diameter inch 0.3736
Fuel Rod Length inch 152.756
Active Fuel Length inch 144.0
Fuel Rod Free Volume inch3 1.30
End Fitting Volume inch3 132.8
Grid Spacer Volume inch3 220.3
Guide/Instrument Tube Volume inch3 71.4
Spring Mass gram/rod 20.0
Fuel Rod Fill Pressure at Manufacture, 20° C psig 500
Moles of Backfill Gas moles/rod 0.031
3.4-85
NAC-STC SAR
Docket No. 71-9235, Revision 17B
Table 3.4-9 Directly Loaded Fuel Basket Component Volumes
Parameter Units
Bottom Weldment inch3
Top Weldment inch3
Support Disks inch3
Spacers inch3
Split Spacers inch3
Threaded Tie Rods inch3
Heat Transfer Disks inch3
Fuel Tubes inch3
Total inch3
3.4-86
June 2018
Revision 18B
Value
2,330
3,565
26,412
1,325
1,896
1,754
20,980
19,185
77,448
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NAC-STC SAR
Docket No. 71-9235
March 2017
Revision 18
How to Specify 0-Rings Denotes
Metallic 0-Rlng
I • Tubing 00 ; I Wall Thlclmeu
{Thirty-Seconcts) ! (Thouundlhs)
I ' II BIIII-IIE • II Materials
1-AlloV 7'18 7 -Stainless 2-Staintea Sleel 304
Steel 32, 8-Stainless 3-Alum-inum St111111316 4-Copper 9-Slainlllss 5-AUC)J 600 Steel 347 6-Allot~750 X-AsSpecified
Exllmpl•: U2312-0362SSEA The &bDve •umple. U2312-0J62~SEA. tnalcaln a type 321 atalnlea!I &IHI Cl-Ring. ~· /2.38 mm) tuba size . . 012 C0.30 mm) wall tt,icl<neu, 3.825' (112.00 mm) 00. aell-energlze<i (ID) ond ,001-.002" (0.03/0.0S mm) ,11,..., c:oaling.
t
Metallic Q.Ring OD {lncnes) (Thauaandths)
'Tyµ. SE-Satl-energized
on ID PF-Prenure filled NP-Not telf
-rgized, not prassure filled
SO--Sttll-ene rgizitd on OD
SX-Sell-ene rgized uspec.
Coatings A-Sil.,., .001 I .002 C0.03/0.051 N--None 8-Sllver .002/ .00310.05/0.08) P-Lead' .001 /.002 (0.03/0.051 D-Tellon• .001 / .003 (0.03/0.08) Fl-Indium .001 / .002 C0.03/0.05) E-Teflon .003/ .004 10.oaro.10) T-Nlckel .0011.002 ro.03/0.05) L-Copper .001 I .002 (0,03/0.05) V-Gold .0005/.001 (0,02/0.03)
X-AaSpec:~d
Fluorocarbon Metallic C-Rings Fluon:>carbc,n Metarnc c-Rings (designated MCA) 11re designed for stllllic: sealing on machinery or equipment and Bre 11vailllble for internal pressure, external preuu~. or axial pressure ID/OD applications. Because C-Rir,gs are desrgned with 11n open side on the pressure side of tile installation. the seal is &elf-energizing. Fluorocarbon C..flinps are offered in round or irregular snapes in a broad range 01 mes trom .126'" (3,2 mm) OD x .Da2" [0,81 mm) fr,11'! ~ight to ov-er 300"' (7620 mm) OD x 2-- (50.80 mm) free neigtlt. They are avail.ltlie in a wide variety of metal alloys and metallic: or Teflon coatinQS. SeaJing application temperature range is trom cryogenic to 3,000" F. 1iesoe C.); p,essur& tolerar,ces are from 10- torr to 100,000 psi (6.804 atm). Where customer requirnments 81'8 large. the C-Ring provides the lowest unit price of any high per1orrnance seal on the market.
•T""""d OuPDM~ ~T-.
@Helimflex Components Division Telephone (803) 783-1880 P.O. Box SB89 FAX 1803) 783-4279 Columbia. South Carolina 29290
4.5-13
NAC-STC SAR
Docket No. 71-9235, Revision l 7C
4.5.2 Blended Polytetrafluoroethylene (PTFE) 0-rings
June2018
Revision 18B
This section contains applicable technical data from a typical manufacturer of blended
polytetrafluoroethylene (PTFE) 0-rings. The PTFE 0-rings are manufactured from virgin
(unreprocessed) polytetrafluoroethylene base material filled with plastic. One product that
satisfies the design requirements is the Fluoroloy K 0-ring manufactured by the Furon Company,
which has an operating temperature range of -450°F to +650°F. NAC has completed
supplemental 0-ring testing and has determined that the operating range of the PTFE 0-rings can
be extended to 735 °F. A description of tests performed and the results are contained in Ce1tified
Test Report D9-3362-1, Applied Technical Services, Inc., February 8, 1989. Another product
that satisfies the design requirements is Parker Compound VM835-75. The compound's
recommended operating temperature range is -40°F to 400 °F.
4.5-14
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NAC-STC SAR
Docket No. 71-9235
March 2017
Revision 18
biasing, the limiting shift scenario corresponds exactly to the position of the fuel assembly in the
cavity, i.e., top dose rates are maximized when the fuel assembly is shifted up and bottom dose
rates are maximized when the fuel assembly is as far down in the cavity as possible. For radial
biasing, the two different shift scenarios are limiting for different transport conditions. The
maximum fuel assembly axial shift is limiting for normal conditions because upper plenum and
upper end fitting hardware move adjacent to the location in the radial shield where the radial lead
shield ends. The limiting shift is downward for accident conditions due to the bottom axial lead
slump, which is adjacent to the lower end fitting hardware source.
The first step in determining limiting PWR dose rates for the directly loaded cask is the
generation of dose rate response functions for generation of minimum cool time tables. For each
array size, at each of 4 burnup, 15 enrichment, and 18 cool time combinations, dose rate profiles
are calculated for both normal and accident transport conditions. Using these dose rate profiles,
the maximum radial dose rates at 2 meters from the railcar are tabulated for normal conditions.
Minimum cool times are calculated to ensure that a decay heat limit of 850 W/assembly is not
exceeded and that the dose rate at 2 meters from the railcar does not exceed 9.5 mrem/hr. The
9 .5 mrem/hr analysis limit was chosen to provide margin against the 10 mrem/hr regulatory limit.
Cool times needed to reach these limits are calculated using linear interpolation on the entire
array of maximum dose rates. The linear interpolation is valid because of the exponential
decrease in source term and, thus, dose rate as a function of time. The interpolated cool time is
rounded up to the next integer year. A sample minimum cool time generation for the 14xl4
reference assembly at 40,000 MWd/MTU is shown in Table 5.4-3. Repeating this analysis for all
fuel types and burnups results in the complete loading table shown in Table 5.4-5. Based on the
loading table, maximum radial dose rates for each fuel type are shown in Table 5.4-4.
The minimum cool times are used to calculate maximum accident condition dose rates at 1 meter
from the cask. The I 000 mrem/hr limit is not exceeded at any of the calculated minimum cool
times.
Based on the radial dose rate results for normal and accident conditions and their application to
the minimum cool time table, the 14x 14 reference assembly provides maximum dose rates. Thus,
top axial and bottom axial response functions have been executed for this assembly only. This
ensures that the maximum axial dose rates for the directly loaded system are captured, and that
variations in burnup, enrichment and minimum cool time are thoroughly examined .
5.4-3
NAC-STC SAR
Docket No. 71-9235, Revision 17C
June 2018
Revision 18B
A summary of the limiting source terms for each transport condition and detector biasing is given
below. All I imiting source terms are taken from the J 4x 14 reference fuel assembly.
Normal Conditions Accident Conditions
Detector Burnup Enrichment Cool Time Burnup Enrichment Cool Time
Biasing [MWd/MTU] [wt% 235U] [Years] [MWd/MTU J [wt% 235UJ [Years]
Radial 40,000 2.3 10 45,000 2.3 14
Top Axial 30,000 2.3 6 45,000 2.3 14
Bottom Axial 40,000 2.3 10 45,000 2.3 14
Three-Dimensional Dose Rates for Directly Loaded Fuel
Further detail on the three-dimensional dose rates are presented in Figures 5 .4-1 through 5 .4-6 for
the limiting 14xl4 reference assembly. Maximum dose rates are tabulated in Tables 5.4-6 and
5.4-7.
The maximum normal conditions surface dose rate is 366 mrem/hr at an axial elevation between
the radial neutron shield and the upper impact limiter. At l meter from the surface of the neutron
shield shell, the maximum dose rate is 20.3 mrem/hr. This dose rate defines the transport index.
The maximum normal conditions dose rate at 2 meters from the cask railcar is 9.5 mrem/hr and
occurs at an axial elevation adjacent to the upper plenum and upper end-fitting elevations. The
maximum accident conditions dose rate at 1 meter from the cask is 665 mrem/hr and occurs at
the cask midplane. The top and bottom axial dose rates are small when compared to the radial
dose rate for the same transport conditions.
Dose rate variations from heat fins in the neutron shield are examined explicitly using azimuthal
detectors that span the entire length of the neutron shield. As shown in Figure 5.4-4, peaks in the
neutron dose rate correspond to dips in the gamma dose rate, and vice versa. Thus, the neutron
dose rate increase resulting from the ducting is offset by the reduction of the gamma dose rate
resulting from the additional shielding provided by the fins. The use of thicker heat fins, which
have a pre-bonded thickness of 8mm to 10mm for 304 stainless steel and 6mm to 8mm for
copper plates, is acceptable as the increased heat fin thickness results in neutron doses that are
within the statistical uncertainty of the existing shielding analysis (i.e.,< 1 o}
Detector descriptions for dose rates on the side of the STC are given in Tables 5.4-8 and 5.4-9 for
normal and accident conditions, respectively. Note that an axial height of 0.0 cm corresponds to
the bottom of the STC cavity.
5.4-4
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NAC-STC SAR June 2018
Revision 18B Docket No. 71-9235, Revision 17B
Table of Contents
8.0 ACCEPTANCE TESTS AND MAINTENANCE PROGRAM ................................. 8-1
8.1 Fabrication Requirements and Acceptance Tests ......................................................... 8.1-1
8.1.1 Weld Procedures, Examination, and Acceptance ............................................. 8.1-1
8 .1.2 Structural and Pressure Tests ............................................................................ 8.1-4
8.1.3 Leakage Tests .................................................................................................... 8.1-6
8.1.4 Co1nponent Tests .............................................................................................. 8.1-9
8.1.5 Tests for Shielding Integrity ........................................................................... 8.1-12
8.1.6 Thermal Test ................................................................................................... 8.1-15
8.1.7 Neutron Absorber Tests for NAC-STC Directly-Loaded Fuel Basket and
for Yankee-MPC and CY-MPC Canistered Fuel Baskets .............................. 8.1-18
8.1.8 Neutron Absorber Tests for MPC-LACBWR Canistered Fuel Basket .......... 8.1-20
8.1.9 Transportable Storage Canister ....................................................................... 8.1-22
8 .1.10 HL W Overpack and Basket ............................................................................ 8 .1-24
8.1.11 Alternative Neutron Absorber/Poison Tests for NAC-STC Directly
Loaded Basket. ................................................................................................ 8.1-27
8.2 Maintenance Program ................................................................................................... 8.2-1
8.2.1 Structural and Pressure Tests of the Casie. ....................................................... 8.2-1
8.2.2 Leak Tests ......................................................................................................... 8.2-2
8.2.3 Subsystems Maintenance .................................................................................. 8.2-3
8.2.4 Valves, Rupture Disks and Gaskets on the Containment Vessel.. .................... 8.2-4
8.2.5 Shielding ........................................................................................................... 8.2-4
8.2.6 Post-Fabrication Thermal Test. ......................................................................... 8.2-4
8.2.7 Miscellaneous ................................................................................................... 8.2-5
8.2.8 Maintenance Program Schedule ....................................................................... 8.2-6
8.3 Quick-Disconnect Valves ............................................................................................. 8.3-1
8.4 Cask Body Fabrication .................................................................................................. 8.4-1
8.4.1 General Fabrication Procedures ........................................................................ 8.4-1
8.4.2 Description of Lead Pour Procedures (Standard Method) ................................ 8.4-5
8.4.3 Description of Lead Pour Procedures (Alternate Method) ............................... 8.4-8
8-i
NAC-STC SAR
Docket No. 71-9235, Revision 17B
List of Figures
June2018
Revision 18B
Figure 8.1-1 Thermal Test Arrangement ............................................................................. 8.1-26
Figure 8.4-1 Typical Arrangement of Lead Pour Equipment (Standard Method) .............. 8.4-12
Figure 8.4-2 Typical Arrangement of Lead Pour Equipment (Alternate Method) .............. 8.4-13
List of Tables
Table 8.1-1 Neutron Absorber Material Minimum 10B Loading (NAC-STC Directly
Loaded Basket) ............................................................................................... 8.1-38
Table 8.1-2 Mechanical Properties of Neutron Absorber (NAC-STC Directly Loaded
Basket) ............................................................................................................ 8.1-38
Table 8.2-1 Maintenance Program Schedule ....................................................................... 8.2-7
8-ii
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NAC-STC SAR
Docket No. 71-9235
March 2017
Revision 18
will be applied in a vertical direction and equally distributed between the two rotation trunnion
recesses by the use of hydraulic rams combined with a load-spreading beam.
Following completion of the rotation trunnion recesses load test, all accessible trunnion recess
welds and load bearing surfaces shall be visually inspected for permanent deformation, galling or
cracking. Inspections utilizing liquid penetrant examination shall be performed in accordance
with the ASME Code, Section V, Article 6. Liquid penetrant acceptance standards shall be as
indicated in paragraph NF-5350 of the ASME Code, Section III, Division I.
Any evidence of permanent deformation, cracking, galling of the load bearing surfaces or
unacceptable dye penetrant results shall be cause for rejection of the rotation trunnion recesses or
related welds.
8.1.2.3 Hydrostatic Testing
A hydrostatic test shall be performed on the NAC-STC cask containment boundary, prior to final
acceptance of the cask, in accordance with the ASME Code, Section III, Division I, Article
NB-6200. The hydrostatic test pressure shall be at least 76 psig, which is 150 percent of the
Maximum Normal Operating Pressure. This test shall be performed in accordance with approved
written procedures. All pressure retaining components, appurtenances, and completed systems
shall be pressure tested.
The vent port will be used for the test connection. Only the vent port quick-disconnect will be
installed during the testing. The hydrostatic test will be performed with the inner lid and the
drain port coverplate installed and torqued.
The hydrostatic test system components, although not part of the cask containment boundary,
will be visually inspected prior to the start of the hydrostatic test. Leakage from the valves or
connections will be corrected prior to the start of the hydrostatic test.
The test pressure gauge installed on the cask will have an upper limit of approximately twice that
of the test pressure. The hydrostatic test pressure shall be maintained for a minimum of 30
minutes, during which time a visual inspection is made to detect any evidence of a leak. Any
evidence of a leak during the minimum hold period will be cause for rejection .
8.1-5
NAC-STC SAR
Docket No. 71-9235, Revision 17C
June 2018
Revision 18B
After completion of the hydrostatic test, the cask containment boundary will be dried and
prepared for visual and/or dye penetrant inspections as appropriate. The components of the cask
containment boundary shall be visually inspected. All accessible welds within the cavity shall be
liquid penetrant inspected. Any evidence of cracking or permanent deformation is cause for
rejection of the affected component.
8.1.2.4 Pneumatic Bubble Testing of the Neutron Shield Tank
A pneumatic bubble test of the neutron shield tank will be performed in accordance with Section
V, A11icle 10, Appendix I, of the ASME Code following final closure welding of the bottom
closure plates. The pneumatic test pressure shall be 12.5 + 1.5/-0 psig, which is 125 percent of
the relief valve set pressure. The test shall be performed in accordance with approved written
procedures.
During the test, the two relief valves on the neutron shield tank will be removed. One of the relief
valves threaded connections will be used for connection of the air pressure line and test pressure
gauge. The other relief valve connection will be plugged with a threaded plug.
•
Following introduction of pressurized air into the neutron shield, a 15-minute minimum soak •
time will be required. Following completion of the soak time, approved soap bubble solution will
be applied to all fin to shell, shell to end plate, and end plate to outer shell welds. The acceptance
criteria for the bubble test will be no air leak from any tested weld as indicated by continuous
bubbling of the solution. If an air leak is indicated, the weld shall be repaired in accordance with
approved weld repair procedures and the pneumatic bubble test shall be repeated until no
unacceptable air leak is observed.
8.1.3 Leakage Tests
Fabrication leakage rate testing is performed on both the NAC-STC transpo11 cask containment
boundary weldment (without the inner lid and inner lid vent and drain port coverplates installed)
during fabrication prior to lead pouring (without the inner lid and inner lid vent and drain port
coverplates installed) upon completion of cask body fabrication (i.e., following lead pouring and
final cask assembly) to demonstrate that the containment boundary weldment, as fabricated, will
provide an appropriate containment capability. The inner lid and with the inner lid vent and
drain port coverplates installed will be leakage rate tested as part of the final fabrication leakage
rate testing per SAR Section 8.1.3.2.
8.1-6 •
• NAC-STC SAR
Docket No. 71-9235, Revision 17C
June 2018
Revision 18B
The leakage rate testing of the NAC-STC containment boundary and closures will be performed
in accordance with the requirements and standards contained in ANSI N 14.5-1997 and the
ASME Code, Section V, Article 10. Helium leakage test procedures shall be developed and
approved by personnel qualified in accordance with the requirements of SNT-TC-1A as a NDT
Level III (Leak Testing) examiner. The leakage rate tests shall confirm that the leakage rate
meets the containment criteria established in Chapter 4 and Table 7.4- 1 ( e.g., leaktight for the
NAC-STC containment boundary weldment; leaktight for Containment Condition A containment
boundary closures with metallic seals for spent fuel transport following long-term storage;
leaktight for Containment Condition BI closures with metallic seals for directly loaded PWR
spent fuel assemblies with burnups of :S 45 GWd/MTU, and for canistered spent fuel and GTCC
waste, and HL W overpacks; leaktight for Containment Condition B3 closures with Viton 0-ring
or metallic containment seals for HBU spent fuel assemblies; or Containment Condition B2 with
Vi ton 0-ring or metallic containment seals with cumulative leakage rate :::; 9 .3 x I 0-5 cm3 /sec
(helium) for containment boundary components with Viton 0-ring seals and leaktight for
metallic containment seals for standard, directly loaded PWR spent fuel assemblies). Leak tests
shall be performed by personnel qualified for helium leakage testing in accordance with the
requirements of ANSI/ASNT CP-189-2006, "Standard for Qualification and Certification of
• Nondestructive Testing Personnel."
8.1.3.1 Containment Boundary Weldment Fabrication Leakage Rate Test
Following the satisfactory completion of hydrostatic pressure testing of the NAC-STC
containment boundary weldment per Section 8.1.2.3, the containment vessel cavity is drained and
cleaned. Per Paragraph 7.3 of ANSI N 14.5-1997, a helium fabrication leakage rate test of the
containment boundary weldment will be performed in accordance with the requirements of
Section V, Article IO of the ASME Code. The containment boundary weldment shall be leakage
tested to demonstrate a leak rate of less than, or equal to, 2x I 0-7 cm3/sec (helium) with a
minimum test sensitivity of Ix 10-7 cm3/sec (helium) to verify the containment boundary
weldment, including containment welds and base materials, is leaktight as defined in ANSI
Nl4.5-1997.
If a leak exceeding the leakage rate acceptance criteria is detected, the affected weld or area of
base metal shall be rejected. Rejected welds or areas of base metal shall be repaired in
accordance with the requirements of Article NB-4450 of the ASME Code. The repaired weld or
base metal area shall be reexamined using the same procedure and acceptance criteria as
specified for the original weld examination. The helium fabrication leakage rate test of the
containment boundary weldment shall then be re-performed in accordance with the original test
• requirements and acceptance criteria prior to final acceptance.
8.1-7
NAC-STC SAR
Docket No. 71-9235
8.1.3.2 Final Fabrication Leakage Rate Testing
March 2017
Revision 18
Upon completion of cask body fabrication, a helium fabrication leakage rate test is performed on
the removable containment boundary closure components and their respective metallic seals or
Viton 0-ring seals (i.e., cask inner lid, lid bolts, inner lid vent and drain port coverplates and
bolts) in accordance with Paragraph 7.3 of ANSI Nl4.5-1997.
• Final containment fabrication leakage rate testing is performed with the cask assembled in
accordance with the cask assembly drawing, except that the vent or drain quick-disconnect is not
installed (Note: The test is repeated to ensure that both the vent and drain port coverplates are
individually leakage tested to the applicable criteria). This ensures that when the cask cavity is
backfilled with helium, helium is present on the containment side of the vent or drain port
coverplate containment 0-ring. Leakage rate tests are performed on the cask lid and the lid pmi
coverplate and their respective 0-ring seals. The test is performed using a helium mass
spectrometer leak detection system by establishing a vacuum of :S 0.1 torr in the seal or 0-ring
annulus of the cask lid and, separately, of the vent and drain port coverplate using the applicable
test port. The cask containment boundary is backfilled with a known concentration of high purity
helium gas. The acceptance criteria for the containment fabrication leakage rate testing of the
removable closures are provided in Section 8.1.3.2.1 and 8.1.3.2.2, depending on the seal •
material and type (e.g., metallic seals or Viton 0-rings). A leakage rate that exceeds the
allowable leakage rate is cause for rejection of the component and seal being tested. Seal
replacement or other corrective actions shall be taken to repair any detected leaks. The
component and replaced seal shall then be retested and re-inspected in accordance with the
original test requirements and acceptance criteria prior to final acceptance. After successful
completion of the leakage tests, quick-disconnects are installed in the inner lid vent and drain
port openings and torqued.
8.1.3.2.1 Metallic Seal Testing Acceptance Criteria
The fabrication leakage rate testing of the containment boundary closures using metallic seals
consists of a series of leak tests. The acceptance criteria for each metallic seal is a detected
leakage rate of :S 2 x 10-7 cm3/sec (helium) at a test system sensitivity of< 1 x 10-7 cm3/sec
(helium) or better.
8.1.3.2.2 Viton 0-Ring Testing Acceptance Criteria
The fabrication leakage rate testing of the containment boundary closures using Viton 0-rings
consists of a series of leak tests. The acceptance criteria for the Viton 0-ring seals is a
8.1-8 •
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NAC-STC SAR
Docket No. 71-9235
March 2017
Revision 18
cumulative (sum of the three individual leakage tests results) detected leakage rate of :'.S 9.3 x 10-5
cm3/sec (helium) at a test system sensitivity of < 4.7 x 10-5 cm3/sec (helium) or better for
standard directly loaded PWR spent fuel assemblies, or leak tight leakage rate of :'.S 2.0 x 10-7
cm3/sec (helium) at a test system sensitivity of< 1.0 x 10-7 cm3/sec (helium) or better for directly
loaded HBU PWR spent fuel assemblies.
8.1.4 Component Tests
Tests performed on individual components are designed to ensure that the component meets the
design requirements for correct and proper operation of the cask system.
Acceptance criteria are established based on the functions and design requirements of the
component being tested.
8.1.4.1 Valves
There are no valves that are part of the NAC-STC containment boundary for transport. Quick
disconnects are installed in the vent, drain and interseal test port openings in the inner lid to
provide access to the cavity, and in the interlid port to provide access to the interlid region.
These fittings serve as valves when the mating parts are connected, and are used to connect
ancillary equipment to the cask cavity for filling, draining, drying, backfilling, gas sampling, and
leak testing operations. Upon removal of the external fitting, the valve in the quick-disconnect
closes automatically. The design and selection of the quick-disconnects is based on similar
equipment and procedures used with other NRC-approved storage and transport casks. For
transport, the quick-disconnects are sealed inside the transport containment boundary using a
bolted coverplate fitted with two 0-ring seals.
There are no rupture disks on the NAC-STC.
Two self-actuating pressure relief valves are installed on the external shell of the neutron shield
to provide for venting of vapor from the shielding material during transport thermal accident
conditions. These valves have stainless steel bodies and an operating pressure range of zero to
200 psig with an adjustable cracking pressure within this range. The cracking pressure is set at
10 psig. These relief valves do not provide a safety function, but have been designed to
minimize recovery effo1is in the unlikely event of a neutron shield overpressure condition .
8.1-9
NAC-STC SAR
Docket No. 71-9235, Revision 17C
8.1.4.2 Gaskets
June 2018
Revision 18B
As described in Section 8.1.3, the containment boundary of the NAC-STC may use either
metallic 0-rings or non-metallic Viton 0-rings. The two 0-ring types require different 0-ring
groove designs and, therefore, may not be used interchangeably and must be used with the inner
lid, vent and drain port coverplates and outer lid having the appropriate 0-ring groove machined
in the component. Metallic 0-rings are required to be used for direct loading of the NAC-STC
with fuel for extended storage and for loading of a transportable storage canister (for transport).
For direct loading of fuel for immediate transpo1i, either metallic or non-metallic 0-rings may be
used.
The outer lid, inner lid, drain port coverplate, vent port coverplate, interlid port cover, pressure
p01i cover, and interseal test plug gaskets are 0-rings. For transport after an extended period of
storage, the containment boundary is formed by the outer metallic 0-ring of the inner lid, the
outer metallic 0-rings on the vent and drain port coverplates, and the interseal test plug metallic
0-rings for the inner lid, the vent port coverplate and the drain port coverplate. The inner
metallic 0-rings of the inner lid, vent port coverplate and drain port coverplate, the metallic
0-ring of the outer lid, and the Viton 0-rings of the interlid and pressure po1i covers provide a
secondary closure to the cask contents. For immediate transport, the containment boundary is
formed by the inner 0-rings of the inner lid and vent and drain port coverplates. A second
boundary is formed by the 0-rings of the outer lid and interseal and pressure port covers.
The 0-ring replacement schedule depends upon the 0-ring material. The metallic 0-ring(s) of
any component shall be replaced prior to reinstallation of the component during loading
operations. Metallic seal replacement is not required prior to the transport of an empty NAC
STC. Those Viton 0-rings that provide the Containment Boundary seal shall be replaced
annually during cask transport operations, or prior to transport if they have been installed longer
than one year. Secondary Boundary (i.e., Non-Containment Boundary) Viton 0-rings shall be
replaced at least once every two years during cask transport operations, or prior to transport if
they have been installed longer than two years.
The containment boundary 0-rings shall be tested and maintained in accordance with the
Maintenance Program Schedule of Table 8.2-1 and the leak test criteria of Section 8.2.2.
8.1-10
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NAC-STC SAR
Docket No. 71-9235, Revision 17C
8.1.4.3 Miscellaneous
June 2018
Revision 18B
The removable transport impact limiters consist of redwood and balsa wood. License drawings
and the supporting analyses specify the crush strengths of the redwood and balsa wood to be
6240 psi± 620 psi and 1550 psi± 150 psi respectively. For manufacturing purposes, verification
of the impact limiter material is accomplished by verifying the densities of the wood. Three
samples from each redwood board are to be tested for density, and the average density of the
samples shall be 23.5 ± 3.5 pounds/cubic foot. Each 15-degree and 30-degree pie shaped section
of the impact limiter shall have a density of 22.3 ± 1.2 pounds/cubic foot in accordance with the
License Drawings. The moisture content for any single redwood board must be greater than 5
percent, but less than 15 percent. The average moisture content for a lot of redwood used in
impact limiter construction must not be greater than 12 percent.
Following final closure welding of the transport impact limiter stainless steel shell, a leak test of
the shell welds shall be performed to verify weld integrity. The following are acceptable test
methods, which may be selected from to verify weld integrity:
1. A test may be performed by evacuating the impact limiter to 75 mbar and performing a
30-minute test to determine if there is any increase in the impact limiter pressure. Any
detected leak shall not exceed 1 x 10-2 cm3/sec. If a leak exceeding this value is detected,
the cause of the leak shall be determined, and the weld repaired and retested.
2. A positive pressure leak test may be performed on each impact limiter to ensure the leak
tightness of the impact limiter shell welds. Remove the test plug and install the necessary
piping to convey oil-free air or gas to the inside of the impact limiter shell. Apply an air
or gas pressure to the inside of the impact limiter shell to initiate the test. Allow the
system to stabilize for at least 15 minutes. Spray all the outside welds with foaming
bubble solution. Examine the limiter welds for indications of continuous bubble
formation. All leaks detected shall be repaired and the leak test re-performed until there
are no leak indications. Upon the completion of leak testing, the test plug shall be
reinstalled.
3. After final closure welding of the transport impact limiter stainless steel shells, a PT
examination may be performed on all shell welds to verify weld integrity. Liquid
penetrant examined per ASME B&PV Section V, Article 6. Acceptance per Section Ill,
Article NF-5350 .
8.1-11
NAC-STC SAR
Docket No. 71-9235, Revisions 17 A and l 7B
8.1.5
8.1.5.1
Tests for Shielding Integrity
Gamma Shield Test
June 2018
Revision 18B
The gamma scan test shall be conducted by continuous scanning or probing over 100 percent of
all accessible cask body surfaces, which directly shield regions where lead was poured, using a
detector and a 6°Co source. Accessible cask surfaces are not only those surfaces that are physically
accessible but also cask surfaces where accurate detector readings can occur. The source strength
shall be of an intensity sufficient to produce a count rate that equals or exceeds three times the
background count rate on the external surfaces of the cask. The count rate shall be maintained
for greater than one minute prior to the start of scanning. The detector scan path spacing ( cask
body exterior surface) will be sufficiently small such that there will be scanning overlap based on
the size detector used and the scanning speed will be 4.5 feet per minute or less. The source scan
path spacing ( cask interior surface) will be on a sized grid pattern that is sufficiently small such
that scanning overlap will occur based on the size detector used.
A gamma scan test is not required for the cask inner closure lid, cask outer closure lid, cask inner
bottom forging, cask outer bottom forging, or cask outer bottom plate. These components shall be
ultrasonic tested in order to demonstrate their soundness as gamma shielding. Ultrasonic testing
shall be performed per ASME B&PV NB-2542.1 using the acceptance standards of Section NB-
2542.2 for forgings and ASME B&PV NB-2532.1 using the acceptance standards of NB-
2532.1 (b) for plates.
The acceptance criteria for the cask body shield test shall be that the shield effectiveness of the
cask body is equal to or greater than the shield effectiveness of a lead and steel mock-up. The
steel thickness of the mockup shall be equivalent to the minimum steel thickness specified on the
License Drawings and the lead thickness shall be equivalent to the minimum lead thickness
specified in the License Drawings Jess 3 percent. The shielding mock-up will be produced using
the same fabrication techniques as those approved for the cask.
Measured count rates that exceed those established by the test mock-up shall cause the
component to be rejected. The rejected areas/components shall be evaluated to determine the
corrective action to be taken. Any repaired areas shall be retested prior to acceptance.
8.1-12
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NAC-STC SAR
Docket No. 71-9235
June 2018
Revision 18B
An additional gamma shield effectiveness test shall be performed on each cask following first
fuel loading. The neutron and gamma shield effectiveness test procedures and acceptance criteria
are described in Section 8.1.5.4.
8.1.5.2 Neutron Shielding Test
The neutron shielding of the NAC-STC is provided by a solid layer ofNS-4-FR, which is a hard
polymer material. A 5.5-inch layer of NS-4-FR is located in the annulus formed by the outer
shell and the 0.236-inch (6 mm) thick neutron shield shell. The neutron shield is divided in
sections by the copper/stainless steel fins. A 2-inch thick layer of NS-4-FR is also installed in
the cask inner lid and in the cask bottom.
The installation of NS-4-FR material in the fabrication of the cask is a special process and, as
such, procedures will be prepared and qualified to ensure that the mix ratios, mixing method,
degassing, pouring, and curing of the material is properly performed. The NS-4-FR raw material
is provided in the form of a 3-part mixing kit. The material content of the raw material is tested
and certified at the time of kit preparation. The neutron shielding material is installed into the
annulus between the outer shell and the neutron shield shell by pouring it with the cask in an
inverted vertical position. Prior to installation, samples from each mix of the actual material
being poured into the annulus are wet density tested to ensure that the material is properly mixed.
Mixes that do not meet the wet density acceptance criteria are rejected. Procedures used for
installation of the material are validated prior to use by destructive examination of a full scale
mock-up of the neutron shield cavity. Qualification of the installation procedure verifies material
homogeneous properties and minimizes the potential deleterious voids.
8.1.5.3 Neutron Shielding Material Testing
The neutron shield prope11ies of NS-4-FR are provided in Chapters I and 3. Each lot (mixed
batch) of neutron shield material shall be tested to verify that the hydrogen concentration, boron
concentration, and neutron shield density meet the requirements specified in Chapters 1 and 3
and the License Drawings. Testing shall be performed by qualified laboratories in accordance
with written and approved procedures. Hydrogen concentration, boron concentration, and
density data for each Jot of neutron shield material shall become part of the quality record
documentation package.
Dimensional inspection of the cavities containing the neutron shielding material shall ensure that
• the required thickness specified in the License Drawings is incorporated into the cask.
8.1-13
NAC-STC SAR
Docket No. 71-9235
June 2018
Revision 18B
The installation of the neutron shielding material shall be performed in accordance with written,
approved, and qualified procedures. The procedures shall ensure that mix ratios and mixing
methods are controlled in order to achieve proper material composition, boron concentration and
distribution, and that pours are controlled in order to prevent gaps or unacceptable voids from
occurring in the material. Procedures shall be qualified by the use of mock-ups to ensure that the
NS-4-FR installation does not result in the creation of unacceptable voids. Wet density data for
each mix of installed neutron shield material shall be maintained as part of the quality record
documentation package.
8.1.5.4 Neutron and Gamma Shield Effectiveness Tests
Following first fuel loading, a neutron and gamma shield effectiveness test shall be performed for
each cask prior to transport. The test shall be performed with the cask loaded with fuel, drained,
vacuum dried and backfilled with helium. The purpose of the test is to document the
effectiveness of the neutron and gamma shielding materials. The test shall be performed in
accordance with detailed, approved written test procedures.
Calibrated neutron and gamma dose rate meters shall be used to measure the neutron and gamma
dose rate at contact with the outer shell of the neutron shield and at 2.3 meters from the surface
(equivalent to 2 meters from the sides of the railcar). Dose measurement points shall be
established on the external surface of the shell at 30° intervals and at five points along the height
of the shield (a total of 60 measuring points). In addition, neutron and gamma dose rate
measurements shall be made of the trunnion areas above the neutron shield, at four points below
the neutron shield, and at the edges and center of the cask top ( outer I id) and cask bottom
surfaces. Dose rates at the top and bottom of the cask shall be measured with the transport
impact limiters installed. The dose rates measured at contact and at 2.3 meters shall be recorded
on the test data sheet, along with the total power of the loaded fuel assemblies; date, time and
location of test; identification and calibration of instrumentation; and identification of test
engineer and operators.
To allow an evaluation of the measured dose rates to be completed, the burn up and cool time for
the actual fuel assemblies loaded into the cask will be determined and recorded. From this fuel
history data, the total actual neutron and gamma source terms will be estimated using ORIGEN
or similar calculations.
8.1-14
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NAC-STC SAR
Docket No. 71-9235
June 2018
Revision 18B
If the measured dose rates exceed the applicable regulatory limits, the licensee shall notify the
NRC. Appropriate corrective measures will be taken, including fuel unloading and correction of
the shielding deficiency. Following corrective actions, the test will be re-performed to the
original acceptance criteria prior to final acceptance.
8.1.6 Thermal Test
Prior to acceptance at the factory, a thermal test shall be performed on each fabricated packaging
to confirm and verify that the fabricated and assembled cask possesses the heat rejection
capabilities predicted by the thermal analyses. The thermal test shall be performed in accordance
with approved written procedures.
8.1.6.1 Thermal Test Set-up
The thermal test set-up 1s shown in Figure 8.1-1 (a). As depicted, the thermal test shall be
performed with the cask positioned horizontally on a test frame. The transport impact limiter or
equivalent insulating material shall be installed on each end of the cask to simulate the transport
configuration. The cask will be located in a covered building in a still environment. The cask
shall be assembled with the basket installed. A thermal test lid with connections for
thermocouple leads and electric heater power cables shall be installed in place of the inner lid.
The outer lid will not be installed for the test. The thermal test lid will be provided with an 0-
ring seal capable of containing the containment cavity helium atmosphere.
Electric heaters shall be installed in each fuel tube. The electric heaters will have an active length
of between 120 and 150 inches and be capable of generating a minimum of 22 kilowatts (kw).
The heaters will be supported in the basket so as to not be in contact with the wall of the fuel
tube. The power supplied to the heater will be recorded throughout the test duration.
Calibrated test thermocouples, with an accuracy of ±2°F, will be installed on the cask basket,
inner shell, and outer neutron shield shell surfaces. The location of the test thermocouples are
shown in Figure 8.1-1. The specific location of the thermocouples are as follows:
TC 1 - basket top steel weldment
TC2 - steel disk at cask basket midpoint
TC3 - aluminum disk at cask basket midpoint
TC4 - basket bottom steel weldment
8.1-15
NAC-STC SAR
Docket No. 71-9235
June 2018
Revision 18B
TC5; TC6; TC7; and TC8 - located at 90° intervals on the inner shell surface at cavity
midpoint
TC9 - top of inner shell surface at 30-40 inches from top of cavity
TC 1 0 - bottom of inner shell surface at 3 0 to 40 inches from base of the cavity
TCI 1; TC12; TC13; and TC14 - located at 90° intervals on the neutron shield shell
surface (at fin tip) at cask midpoint
TC15 - top of neutron shield shell surface (at fin tip) at 30-40 inches from top of neutron
shell
TC16-bottom of neutron shield shell surface (at fin tip) at 30-40 inches from bottom of
neutron shield shell
TC 1 7 - top of upper forging
TC 18 - outer shell surface at centerline of cask bottom fac;:e
TC 19 - inner fuel tube wall surface near the center of the cask basket
TC20 - ambient temperature of testing area
The output of the test thermocouples will be recorded throughout the test by a strip chart
recorder.
8.1.6.2 Test Procedure
With the cask assembled and instrumented as described above, the cask cavity is evacuated and
backfilled to 1.0 atmosphere absolute (14.6 psia) with helium. Power will be applied to the
heaters to simulate the cask contents. After initiation of power to the heaters, the temperatures of
all thermocouples and heater power levels will be monitored and recorded on data sheets at 60
minute intervals. Power will be maintained to the electrical heaters until the cask has reached
thermal equilibrium.
For the purpose of the test, thermal equilibrium is defined as being achieved when over two
consecutive hours:
~tTCJ3::::; 2°F/hr, and
~tTc3::::; 2°F/hr
Based upon the thermal heat-up evaluation, thermal equilibrium should be achieved m
approximately five days.
8.1-] 6
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NAC-STC SAR
Docket No. 71-9235
June 2018
Revision 18B
After verification of thermal equilibrium, final temperature measurements will be recorded for all
test thermocouples. The final power readings for the electric heaters will also be recorded. The
strip chart will be marked to indicate the time of the final cask measurements. The printout of the
strip chart recorder and the completed test data sheets will be incorporated into an approved final
thermal test report. The test will be determined to be acceptable if the acceptance criteria of
Section 8.1.6.3 are met.
If the acceptance criteria are not met, the cask will not be accepted until appropriate corrective
actions are completed. Upon completion of corrective actions, the cask shall be retested to the
original test requirements and acceptance criteria.
8.1.6.3 Acceptance Criteria
The purpose of the thermal test is to confirm the heat rejection capabilities of the as-built cask
are acceptable and correspond to the temperatures calculated by thermal analyses for the directly
loaded (uncanistered) configuration presented in Chapter 3.0 of this application .
Package heat dissipation acceptance testing assures: 1) maximum material temperatures do not
exceed material allowables; and that 2) measured temperature gradients are less than the thermal
gradients calculated in the package thermal analyses.
The thermal acceptance test is accepted when the following criteria are met:
I) When corrected for physical test boundary conditions and heat load, the following
. measured temperatures are not exceeded:
TC No. Location
TCl Top Basket Steel Weldment
TC3
TC2
TC4
TC5-TC8
TCl l-TC14
TC17
TC18
TC19
Aluminum Disk Center
Steel Support Disk Center
Basket Bottom Steel Weldment
Cask Inner Shell
Neutron Shield Shell
Cask Top Forging
Cask Bottom
Tube Wall
8.1-17
Temperature °F
435
485
495
475
330
240
200
330
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NAC-STC SAR
Docket No. 71-9235
June 2018
Revision 18B
2) The measured temperature gradient across the central steel disk from TC2 to the
average of TC5, TC6; TC7 and TC8 is less than 200°F;
3) The measured temperature gradient across the central aluminum disk from TC3 to
the average of TC5; TC6; TC7 and TC8 is less than l 90°F; and
4) The measured temperature gradient across the cask body as measured by
thermocouple pairs TC5-TC13; TC6-TC14; TC7-TC11; and TC8-TC12 are less
than 90°F.
8.1.7 Neutron Absorber Tests for NAC-STC Directly-Loaded Fuel Basket and for
Yankee-MPC and CY-MPC Canistered Fuel Baskets
Two alternate neutron poison materials, BORAL and Ta!Bor, have been qualified by NAC for
use in the NAC-STC directly loaded, the Yankee-MPC and the CY-MPC fuel baskets. For the
NAC-STC directly loaded basket BORAL acceptance testing is described in this section (Section
8.1.7) while a generic metal matrix composite (aluminum based) and borated aluminum
acceptance/qua! ification program is described in Section 8.1 .11. The generic program is
designed to demonstrate structural, thermal, and nuclear requirements are met without
•
specification of a particular manufacturer or material. Ta!Bor, an MMC material, is excluded •
from the generic acceptance/qualification program, and is included in Section 8.1.7.
BORAL is manufactured by Ceradyne Corporation, Chicoutimi (Quebec), Canada under a
Quality Assurance/Quality Control program in conformance with the requirements of 10 CFR
50, Appendix B. The manufacturing process consists of several steps: the first step is the mixing
of the aluminum and boron-carbide powders that form the core of the finished material, with the
amount of each powder a function of the desired 10B areal density. The methods used to control
the weight and blend of the powders are patented and proprietary processes of AAR Advanced
Structures (AAR) (subsequently Ceradyne). The mixture of powders is placed in an aluminum
box with walls approximately one inch thick. The top lid is welded in place. This "ingot" is
heated for several hours and then is hot-rolled to produce the sheet of design thickness. The
rolling process densifies and bonds the powder mixture. The aluminum box walls become the
cladding for the Al-B4C core.
Ta!Bor is manufactured by Talon Composites, Inc. (Ta!Bor was formerly called Boralyn, and
was produced by Alyn Corporation. Alyn Corporation went out of business and Talon
Composites acquired the major production equipment and the patent rights for Boralyn. Ta!Bor
is essentially identical to Boralyn.) Ta!Bor is manufactured and controlled using a Quality
Assurance program that is compliant with the applicable requirements of 10 CFR 50, Appendix,
8.1-18 •
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Docket No. 71-9235
June 2018
Revision 18B
B. TalBor is a metal matrix composite (MMC). The aluminum and B4C powders are mixed to
the specified JOB areal density and the powder mixture is vacuum sintered and hot pressed to
achieve a fully dense billet. The billet is extruded, then cut and rolled to the design thickness.
After manufacturing, test samples from each batch of neutron absorber (poison) sheets shall be
tested to verify the presence, proper distribution, and minimum weight percent of JOB. Neutron
transmission testing or augmented wet chemistry testing may be used. The tests shall be
performed in accordance with approved written procedures.
8.1.7.1 Neutron Absorber Material Sampling Plan
The neutron absorber sampling plan is selected to demonstrate a 95/95 (95% probability and 95%
confidence level) statistical confidence level in the neutron absorber sheet material compliance
with the specification. In addition to the specified sampling plan, each sheet of material is
visually and dimensionally inspected using at least 6 measurements (along the edges near each
corner and the longitudinal centerline) on each sheet. No rejected neutron absorber sheet is used.
The sampling plan is supp01ied by written and approved procedures .
The sampling plan requires that a coupon sample be taken from each sheet of the first set of 100
sheets of absorber material. Thereafter, coupon samples are taken from 20 randomly selected
sheets from each set of 100 sheets. This 1 in 5 sampling plan continues until there is a change in
lot or batch of constituent materials of the sheet (i.e., boron carbide powder, aluminum powder,
or aluminum extrusion), or a process change, at which time the sampling process is reinitiated as
previously described. The sheet samples are indelibly marked and recorded for identification.
This identification is used to document neutron absorber test results, which become part of the
quality record documentation package.
8.1.7.2 Wet Chemistry Test Performance
An approved facility with chemical analysis capability shall be selected to perform the wet
chemistry tests. The tests will ensure the presence of boron and enable the calculation of the JOB
areal density. Acceptability of the uniformity of boron distribution is based on the
manufacturer's material qualification tests.
The most common method of verifying the acceptabi 1 ity of neutron absorber material is the wet
chemistry method-a chemical analysis where the aluminum is separated from a sample with
known thickness and volume. The remaining boron-carbide material is weighed and the areal
8.1-19
NAC-STC SAR
Docket No. 71-9235
June 2018
Revision 18B
density of 10B is computed. A statistical conclusion about the BORAL or Ta!Bor sheet from
which the sample was taken and that batch of sheets may then be drawn based on the test results
and the established manufacturing processes previously noted.
BORAL and Talbor sheets are required to contain a minimum 0.020 g 10B/cm2 which is credited
at 75% effectiveness in Chapter 6.
8.1.7.3 Neutron Absorption Transmission Test Performance
An approved facility with a neutron source and neutron detection capability shall be selected to
perform the described tests, if the neutron absorption transmission test method is used. The tests
will assure that the neutron absorption capacity of the material tested is equal to, or higher than,
the given reference value and will verify the uniformity of boron distribution. The principle of
measurement of neutron absorption is that the presence of boron results in a reduction of neutron
flux between the thermalized neutron source and the neutron detector-depending on the
material thickness and boron content.
Typical test equipment will consist of thermal neutron source equipment, a neutron detector and
a counting instrument. The test equipment is calibrated using a known standard, whose 10B
content has been checked and verified by an independent method such as chemical analysis. This
calibration process shall be repeated daily (every 24 hours) while tests are being performed.
8.1.7.4 Acceptance Criteria
The neutron transmission test results shall be considered acceptable if the minimum 10B areal
density is determined to be equal to, or greater than, that specified on the fuel tube drawings.
Any specimen not meeting the acceptance criteria shall be rejected and all of the sheets from that
batch shall be similarly rejected unless coupons from each individual absorber plate are tested
and confirmed to meet or exceed the specified areal density.
8.1.8 Neutron Absorber Tests for MPC-LACBWR Canistered Fuel Basket
Neutron absorber material ( commercially available as BORAL ®), in the form of sheets consisting
of boron-carbide evenly dispersed within a matrix of aluminum and clad with aluminum, is used
in the NAC-MPC transportable storage canister fuel baskets. The manufacturing process consists
of several steps - the first being the mixing of the aluminum and boron-carbide powders that
form the core of the finished material, with the amount of each powder a function of the desired
8.1-20
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Docket No. 71-9235
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108 areal density. The methods used to control the weight and to blend the powders were
patented and proprietary processes of AAR and, subsequently, of Ceradyne Corporation of
Chicoutimi (Quebec), Canada.
After manufacturing, test samples from each batch of BORAL ® neutron absorber (poison) sheets
shall be tested using wet chemistry or neutron absorption techniques to verify the presence,
proper distribution, and minimum weight percent of 10B. The tests shall be performed in
accordance with approved written procedures.
8.1.8.1 Neutron Absorber Material Sampling Plan
The neutron absorber sampling plan is selected to demonstrate a 95/95 (95% probability and 95%
confidence level) statistical confidence level in the neutron absorber sheet material
compliance with the specification. In addition to the specified sampling plan, each sheet of
material is visually and dimensionally inspected using at least six measurements (along the edges
near each corner and the longitudinal centerline) on each sheet. No rejected neutron absorber
sheet is used. The sampling plan is supported by written and approved procedures.
The sampling plan requires that a coupon sample be taken from each sheet of the first set of 50
sheets of absorber material. Thereafter, coupon samples are taken from 10 randomly selected
sheets from each set of 50 sheets. This I in 5 sampling plan continues until there is a change in
lot or batch of constituent materials of the sheet (i.e., boron carbide powder, aluminum powder,
or aluminum extrusion), or a process change, at which time the sampling process is reinitiated as
previously described. The sheet samples are indelibly marked and recorded for identification.
This identification is used to document neutron absorber test results, which become part of the
quality record documentation package.
8.1.8.2 Wet Chemistry Test Performance
An approved facility with chemical analysis capability shall be selected to perform the wet
chemistry tests. The tests will ensure the presence of boron and enable the calculation of the 10B
areal density. Acceptability of the uniformity of boron distribution is based on the
manufacturer's material qua! ification tests.
The most common method of verifying the acceptability of neutron absorber material is the wet
chemistry method - a chemical analysis where the aluminum is separated from a sample with
known thickness and volume. The remaining boron-carbide material is weighed and the areal
• density of 108 is computed. A statistical conclusion about the BORAL ® sheet from which the
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Docket No. 71-9235
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Revision 18B
sample was taken and that batch of BORAL ® sheets may then be drawn based on the test results
and the established manufacturing processes previously noted.
8.1 .8.3 Neutron Absorption Test Performance
An approved facility with a neutron source and neutron detection capability shall be selected to
perform the described tests, if the neutron absorption test method is used. The tests will assure
that the neutron absorption capacity of the material tested is equal to, or higher than, the given
reference value and will verify the uniformity of boron distribution. The principle of
measurement of neutron absorption is that the presence of boron results in a reduction of neutron
flux between the thermalized neutron source and the neutron detector-depending on the
material thickness and boron content.
Typical test equipment will consist of thermal neutron source equipment, a neutron detector and
a counting instrument. The test equipment is calibrated using standards whose 10B content has
been checked and verified by an independent method such as chemical analysis. The highest
permissible counting rate is determined from the neutron counting rates of the reference sheet(s),
•
which should be ground to the minimum allowable plate thickness. This calibration process shall •
be repeated daily (every 24 hours) while tests are being performed.
8.1.8.4 Acceptance Criteria
The wet chemistry test results shall be considered acceptable if the 10B areal density is
determined to be equal to, or greater than, that specified on the fuel tube drawings. The neutron
absorption test shall be considered acceptable if the neutron count determined for each test
specimen is less than or equal to the highest permissible neutron count rate determined from the
BORAL standard, which is based on the 10B areal density specified on the fuel tube drawings.
Any specimen not meeting the acceptance criteria for either test method shall be considered to be
nonconforming material and shall be evaluated within the NAC International QA Program.
Nonconforming material shall be assigned one of the following dispositions: "use-as-is,"
"rework" or "reject." Only material that is determined to meet all applicable conditions of the
license will be accepted.
8.1.9 Transportable Storage Canister
The transportable storage canister is constructed of Type 304L (Yankee-MPC and CY-MPC) or
304/304L (MPC-LACBWR) stainless steel and is fabricated by welding. If circumferential •
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Docket No. 71-9235
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Revision 18B
welds are required to join two shell sections, the seam welds shall not be aligned within 45°
circumferentially. The welded cylinder is closed at the bottom by a circular plate welded to the
shell wall. The top of the cylinder is closed by two field-installed circular plates, welded to the
canister shell wall following fuel loading.
The transportable storage canister is a welded closed component. The canister serves as the
confinement boundary component of the NAC-MPC System during storage of spent fuel in the
vertical concrete cask.
The finished surfaces of all canister welds are visually examined in accordance with ASME Code
Section V, Article 9, to verify that the components are assembled in accordance with the License
Drawings and that the components are free of nicks, gouges, and other damage. The acceptance
criteria for the visually examined welds for the Yankee-MPC and the CY-MPC canisters is in
accordance with ASME Code Section VIII, Division 1, UW-35 and UW-36 and Section III,
Subsection NB, NB-4424 and NB-4427. The acceptance criteria for the visually examined welds
of the MPC-LACBWR canister are in accordance with ASME Code, Section IIl, Subsection NF,
NF-5360 .
The seam and girth welds in the transpo1iable storage canister shell are full-penetration welds
that are radiographic (RT) examined in accordance with ASME Code Section V, Article 2. The
acceptance criteria for the RT-examined welds is that specified in ASME Code Section III,
Subsection NB, Article NB-5320. The canister shell to bottom plate weld is a full-penetration
double-bevel weld with an inside fillet weld that is ultrasonic examined in accordance with
ASME Code Section V, Article 5, with acceptance criteria as specified in ASME Code Section
III, Subsection NB, Article NB-5330. The final surfaces of the seam and girth welds in the
canister and the canister shell to bottom plate weld are also liquid penetrant examined in
accordance with ASME Code Section V, Article 6, with the acceptance criteria being that
specified in ASME Code Section III, Subsection NB, Article NB-5350.
Field installed partial-penetration groove welds attach the shield and structural lids (Yankee
MPC and CY-MPC) or the closure lid (MPC-LACBWR) to the canister shell, and the vent and
the drain port covers to the shield lid (Yankee-MPC and CY-MPC) or the inner and outer vent
and drain port covers to the closure lid (MPC-LACBWR) after the canister is loaded. The
closure ring for the MPC-LACBWR canister is welded to both the canister shell and closure lid
by partial penetration welds. For the Yankee-MPC and CY-MPC canister, the root and final
• surfaces of the shield lid weld are liquid penetrant examined. For the MPC-LACBWR canister,
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NAC-STC SAR
Docket No. 71-9235
June 2018
Revision 18B
the closure lid to canister shell weld is progressively liquid penetrant examined at the root, mid
plane and final surfaces. The structural lid to shell weld for the Yankee-MPC and CY-MPC
canisters is progressively liquid penetrant examined at the root, every 3/8-inch weld layer and
final surface. Canister vent and drain port cover welds are liquid penetrant examined at the root
and final surfaces unless the welds are completed in a single pass. Welds completed in a single
pass require only final surface liquid penetrant examination.
All liquid penetrant examinations are completed in accordance with ASME Code, Section V,
Article 6. Acceptance criteria for all liquid penetrant examinations are as specified in ASME
Code Section lII, Division 1, Subsection NB, Article NB-5350.
The Yankee-MPC and CY-MPC canister shield lid welds are helium leakage tested in
accordance with ASME Code Section V, Article 10, Appendix V, using a minimum leak rate test
sensitivity of 1 x 10-7 cm3/sec (helium). The MPC-LACBWR canister closure lid to canister
shell weld is hydrostatically tested following completion of the weld.
The fabricator of the transportable storage canister will establish a written weld inspection plan
in accordance with an approved quality assurance program. The weld inspection plan will
include visual, liquid penetrant, ultrasonic, and radiographic examination. In addition, the weld
inspection plan will identify the welds to be examined, the sequence of the examinations, the
type of examination method to be used, and the criteria for acceptance of the weld in accordance
with the applicable sections of the ASME Code.
8.1.10 HL W Overpack and Basket
The HL W Overpack is constructed of Type 304/304L stainless steel and is fabricated by
welding. If circumferential welds are required to join two shell sections, the seam welds shall
not be aligned within 45° circumferentially. The welded cylinder is closed at the bottom by a
circular plate welded to the shell wall. The top of the cylinder is closed by a field-installed
circular plate, welded to the canister shell wall following HL W canister loading.
The HL W Overpack is a welded closed component.
The finished surfaces of all HL W Overpack welds are visually examined in accordance with
ASME Code Section V, Articles 1 and 9, to verify that the components are assembled in
accordance with the License Drawings and that the components are free of nicks, gouges, and
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Docket No. 71-9235
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Revision 18B
other damage. The acceptance criteria for the visually examined welds for the HL W overpack
are in accordance with ASME Code Section VIII, Division 2, Section 7.5.2.2.
The seam, girth, and shell to bottom plate welds in the HL W overpack shell are full-penetration
welds that are dye penetrant (PT) examined in accordance with ASME Code Section V, Articles
1 and 6. The acceptance criteria for the PT examined welds are those specified in ASME Code
Section VIII, Division 2, Section 7.5.7.2.
Field installed partial-penetration groove welds attach the closure lid to the HL W Overpack shell
after HL W canister loading. The closure lid to canister shell weld is visually examined at the
final surface. Visual examinations are completed in accordance with ASME Code, Section V,
Articles 1 and 9. Acceptance criteria for all visual examinations are as specified in ASME Code
Section VIII, Division 2, Section 7.5.2.2.
The HL W Overpack basket is fabricated from Type 304 stainless steel and is fabricated by
welding. If circumferential welds are required to join two shell sections, the seam welds shall
not be aligned within 45° circumferentially. The five (5) welded HL W cylinder cells are closed
at the bottom by a plate welded to the cell wall where accessible. The top of the HL W cy I ind er
cell is open to allow vertical dry loading of a HLW canister.
The finished surfaces of all HL W overpack basket assembly welds are visually examined in
accordance with ASME Code Section V, Articles 1 and 9 to verify that the components are
assembled in accordance with the License Drawings and that the components are free of nicks,
gouges, and other damage. The acceptance criteria for the visually examined welds for the HL W
overpack basket are in accordance with ASME Code Section III, Subsection NF, NF-5360.
Liquid penetrant (PT) examination will be performed on all HL W Overpack basket welds in
accordance with ASME Code, Section V, Articles 1 and 6. Acceptance criteria for the PT
examined welds shall be in accordance with ASME Code, Section III, Subsection NF, NF-5350.
The fabricator of the HLW Overpack and basket assemblies will establish a written weld
inspection plan in accordance with an approved quality assurance program. The weld inspection
plan will include visual and liquid penetrant examination. In addition, the weld inspection plan
will identify the welds to be examined, the sequence of the examinations, the type of
examination method to be used, and the criteria for acceptance of the weld in accordance with
the applicable sections of the ASME Code .
8.1-25
NAC-STC SAR
Docket No. 71-9235
Figure 8.1-1 Thermal Test Arrangement
I I I
le-,-;=,=-· -1--- - - -l I I
.__~POIIER-=~=Pl'L-Y-------~ 1D HEAlI!IS (2e) I
I I I I L
lC12i\ r "-M '11:11
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-, I I I
I I
I I I I I _J
~~---~---~----~~ ---- - -- - --- - - - -----....::
March 2017
Revision 18
IIIP~T Wll'ERS OR [l;UVN.Dff IN~TI~
TEST SET-UP AND ........_"JEST 111M€
(a) EXTERNAL THERMOCOUPLE LOCATIONS
r=
\ . .. • -=
• , ~
1•.,
DISK PLAN VIEW mrn~
(b) INTERNAL CAVITY AND BASKET THERMOCOUPLE LOCATIONS
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NAC-STC SAR
Docket No. 71-9235
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Revision 18
During periods when the cask is not in use for transport, the periodic leakage rate test need not
be performed on an annual basis, but shall be re-performed prior to returning the cask to service
and use as a transport package.
8.2.2.3 Periodic and Maintenance Leakage Test Acceptance Criteria
8.2.2.3.1 Metallic Seal Testing Acceptance Criteria
The periodic or maintenance leakage testing of the containment boundary closures using metallic
seals consists of a series of leak tests. The acceptance criteria for each metallic seal is a detected
leakage rate of ::S 2 x 10-7 cm3/sec (helium) at a test system sensitivity of< 1 x 10-7 cm3/sec
(helium) or better.
Unacceptable leakage test results shall be cause for rejection of the component tested. Corrective
actions, including repair or replacement of the seals and/or closure component, shall be taken and
documented as appropriate. The leakage test shall be repeated and accepted prior to returning
the cask to service .
8.2.2.3.2 Viton 0-Ring Testing Acceptance Criteria
The periodic or maintenance leakage testing of the containment boundary closures using Viton
0-rings consists of a series of leak tests. The acceptance criteria for the Viton 0-ring seals is a
cumulative (sum of the three individual leakage tests results) detected leakage rate of ::S 9.3 x 10-5
cm3/sec (helium) at a test system sensitivity of< 4.7 x I 0-5 cm3/sec (helium) or better for
standard directly loaded PWR spent fuel assemblies; or leak tight leakage rate of ::S 2.0 x I 0-7
cm3/sec (helium) at a test system sensitivity of< I .0 x 10-7 cm3/sec (helium) or better for directly
loaded HBU PWR spent fuel assemblies.
Unacceptable leakage rate test results shall be cause for rejection of the component tested.
Corrective actions, including repair or replacement of the 0-rings and/or closure component,
shall be taken and documented as appropriate. The leak test shall be repeated and accepted prior
to returning the cask to service.
8.2.3 Subsystems Maintenance
There are no subsystems maintenance requirements on the NAC-STC.
8.2-3
NAC-STC SAR
Docket No. 71-9235, Revision l 7B
8.2.4 Valves. Rupture Disks and Gaskets on the Containment Vessel
June2018
Revision l 8B
There are no valves on the NAC-STC packaging providing a containment function. Four quick
disconnects, one each on the vent, drain, inner lid interseal test and interlid ports, are provided
for ease of cask operation.
The quick-disconnect shall be inspected during each cask loading and unloading operation for
proper performance and function. As necessary, the subject quick-disconnect shall be replaced.
The quick-disconnects shall be replaced every two years during transport operations, and
following fuel unloading after extended storage.
There are no rupture disks on the NAC-STC containment vessel.
All 0-rings on the NAC-STC shall be visually inspected for damage during each cask operation.
All metallic 0-rings shall be replaced during each cask loading sequence. Viton 0-rings shall be
replaced annually and as required, based on leak testing results and inspections during
operations. PTFE 0-rings shall be replaced if damage is noted during the visual inspection and
every two years during transport operations.
8.2.5 Shielding
The gamma and neutron shields of the NAC-STC packaging do not degrade with time or usage.
The radiation surveys performed by licensees prior to transport and upon receipt of the loaded
cask provide a continuing validation of the shield effectiveness of the NAC-STC.
8.2.6 Post-Fabrication Thermal Test
Prior to acceptance at the factory, the heat rejection capability of each fabricated NAC-STC
packaging has been confirmed and verified by the thermal test as described in Section 8.1.6.
Prior to each fuel loading in accordance with the operating procedures of Chapter 7, a visual
inspection on the cask including the visual inspection on the radial neutron shield shell (see
Section 8.2.7) will confirm that there is no change of the heat transfer capability of the
packaging. However, a post-fabrication thermal test shall be performed on an operational NAC
STC packaging if, during handling or transport operations, the packaging experiences an adverse
event such as fire, drops or impacts that result in obvious damage to the neutron shield. The
8.2-4
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Docket No. 71-9235, Revision 17B
June 2018
Revision 18B
post-fabrication thermal test shall be performed in accordance with the fabrication thermal
acceptance test as described in Section 8.1.6 with approved written procedures including defined
acceptance criteria. The packaging shall meet the acceptance criteria of the post-fabrication
thermal test prior to a return to radioactive material transpo11 operations.
8.2.7 Miscellaneous
The transport impact limiters shall be visually inspected prior to each shipment. The limiters
shall be visually inspected for gross damage or cracking to the stainless steel shells in accordance
with approved written procedures and established acceptance criteria. Impact limiters not
meeting the established acceptance criteria shall be rejected until repairs are performed and the
component re-inspected and accepted.
The cask cavity shall be visually inspected prior to each fuel loading. Evidence of gross scoring
of the cavity surface, or build-up of other foreign matter in the cask cavity that could block the
cavity drainage paths shall be cause for rejection of the cask for use until approved maintenance
and/or repair activities have been acceptably completed. The basket assembly for the directly
loaded (uncanistered) or canistered configuration (prior to initial loading) shall be visually
inspected for deformation of the basket disks or tubes. Evidence of damage shall be cause for
rejection of the basket until approved repair activities have been completed, and the basket has
been re-inspected and approved for use.
The radial neutron shield shell shall be visually inspected prior to each fuel loading. Any crack,
gauge, or gross deformation that could indicate damage of the heat transfer fins shall be cause for
rejection of the cask for use until approved maintenance and/or repair activities have been
acceptably completed.
The overall condition of the cask, including the fit and function of all removable components,
shall be visually inspected and documented during each cask use. Components or cask conditions
which are not in compliance with the Certificate of Compliance shall cause the cask to be
rejected for transport use until repairs and/or replacement of the cask or component are
performed, and the component re-inspected and accepted.
The results of the visual inspections, leakage tests, shielding and radiological contamination
surveys; fuel identification information for the package contents; date, time, and location of the
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NAC-STC SAR
Docket No. 71-9235, Revision l 7B
June 2018
Revision l 8B
cask loading operations; and remarks regarding replaced components shall be included in the
cask loading report for each loaded cask transport. The requirements of the cask loading report
shall be detailed in the NAC-STC Operations Manual.
8.2.8 Maintenance Program Schedule
Table 8.2-1 presents the overall maintenance program schedule for the NAC-STC.
8.2-6
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Docket No. 71-9235, Revision 17B
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Revision 18B
Table 8.2-1 Maintenance Program Schedule
Task Frequency Cavity Visual Inspection Prior to Fuel Loading Basket Visual Inspection Prior to Fuel Loading 0-ring Visual Inspection Prior to Fuel Loading Outer Lid, Inner Lid and Port Coverplate
Bolt Visual Inspection Prior to installation during each use Radial Neutron Shield Shell Visual Prior to Fuel Loading Inspection Cask Visual
and Proper Function Inspections Prior to each Shipment Lifting and Rotation Trunnion
Visual Inspection Prior to each Shipment Liquid Penetrant Inspection of surfaces Annually during use and accessible welds
Maintenance Periodic Leakage Rate Test For Viton 0-rings, annually or when replaced. of Inner Lid and Port Coverplate 0- For metallic 0-rings, prior to each loaded transport.
rings Preshipment Leakage Rate Test Prior to loaded transport for casks with Viton 0-
rings Transport Impact Limiter Visual Prior to each shipment Inspection Quick-disconnect
Inspection for Proper Function During each Cask Loading/Unloading Operation Quick-disconnect Replacement Every two years during transport operations Metallic 0-ring Replacement Prior to installation for a loaded transport Viton 0-ring Replacement Annually, or more often, based on inspections
during use or leakage test results Inner and Outer Lid Bolt Replacement Every 240 bolting cycles
(Every 20 years at 12 cycles per year) PTFE 0-ring Replacement Every two years during transport operations or as
required by inspection Periodic Leakage Rate Test Performed within 12 months prior to each shipment
for containment boundary Viton 0-rings. No testing needed for out-of-service packaging or for casks provided with containment boundary metallic seals as metallic seals are replaced and maintenance leakage tested during each loading operation.
Post-Fabrication Thermal Test Performed after a cask experiences an adverse event such as fire, drops or impacts that result in obvious damage to the neutron shield. The cask shall pass the pre-fabrication thermal test prior to being used in a subsequent fuel transport.
8.2-7
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NAC-STC SAR
Docket No. 71-9235, Revision 17C
8.4 Cask Body Fabrication ·
8.4.1 General Fabrication Procedures
June 2018
Revision 18B
The NAC-STC cask body is a welded structure of stainless steel plates and forgings. Chemical
Copper lead is poured in place between the inner and outer shells to serve as the main gamma
shielding material. NS-4-FR is poured in place between the neutron shield shell and the outer
shell. NS-4-FR is also form fit between the bottom inner forging and the bottom plate and in the
inner lid. Welding on the NAC-STC shall be performed in accordance with the requirements of
the ASME Code and the American Welding Society (A WS) Structural Welding Code - Steel
(ANSl/A WS D.1-1) as specified on the NAC-STC License Drawings and Section 8.1.1.
The general fabrication procedures for the NAC-STC are summarized, herein, to facilitate an
understanding of the component configurations and the weld locations shown on the license
drawings.
Each of the two inner shell rings (upper and lower) is rolled from Type XM-19 stainless steel
plate and seam welded longitudinally. The outside diameter of each inner shell ring is machined
to the defined transition section dimensions. The minimum length of each Type XM-19 shell ring
shall be in accordance with the License Drawings. The central inner shell sections are each
rolled from Type 304 stainless steel plate and seam welded longitudinally. The number and
length of the individual inner shell sections to be used to obtain the required total inner shell
length is optional. The inner shell sections are girth welded to each other and the inner shell rings
are girth welded on each end of the inner shell. Longitudinal seam welds in adjacent inner shell
sections shall be offset at a minimum of 15 degrees for girth-welded sections.
After initial rough machining and final weld preparation, the top forging and the bottom inner
forging are individually welded to the opposite ends of the inner shell/inner shell ring weldment
to form the cask cavity. The preparation, examination, and acceptance procedures for the welds
are described in Section 8.1.1 and defined on the License Drawings. Following inspection and
acceptance of the welds, the top forging and the outside diameter of the cask cavity weldment are
final machined. Following final machining of both sides of the inner shell, an ultrasonic
thickness test of the inner shell wall of the cask cavity shall be performed to confirm that the wall
thickness of any location on the shell is not Jess than 1.46 inches (37.1 mm). A wall thickness at
any location of less that 1.46 inches (37.1 mm) will be cause for rejection. Rejected areas of the
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NAC-STC SAR
Docket No. 71-9235, Revision 17C
June 2018
Revision 18B
shell wall can be repaired by weld overlay using approved written weld overlay procedures.
Following repair, the repaired areas shall be examined in accordance with the original inspection
requirements and acceptance criteria.
Following thickness testing, the cask cavity weldment, which is the NAC-STC primary
containment boundary, shall be hydrostatically tested according to ASME Code, Section III,
Subsection NB-6000, as described in Section 8.1.2.3. The cask cavity weldment is dried, the
primary containment boundary welds are liquid penetrant examined in accordance with ASME
Code, Section V, Article 6, and the welds are accepted in accordance with ASME Code, Section
III, Subsection NB-5350. The cask cavity weldment is then helium leak tested to verify that the
Containment System Fabrication Verification leak rate is satisfied, as described in Section 8.1.3.
Each of the outer shell sections is rolled from Type 304 stainless steel plate and seam welded
longitudinally. The number and length of outer shell sections to be used to achieve the required
total outer shell length is optional. The outer shell sections are girth welded to each other and the
inside diameter of the "outer shell weldment" is final machined. Longitudinal seam welds in
adjacent outer shell sections are offset at a minimum of 15 degrees for gi1ih-welded sections .
The outer shell weldment is welded to the cask cavity weldment at the top forging/outer shell
interface to form the "body weldment." The preparation, examination, and acceptance
procedures for the welds are described in Section 8.1.1 and defined on the License Drawings.
The body weldment is inverted (closure end down) in a pit or other sheltered location m
preparation for lead pouring. A temporary dam extension and supports are welded to the open
end of the outer shell to permit the full length of the lead shell to be poured and to maintain the
outer shell position. "Backing bars" are tack-welded on the inside diameter of the outer shell
overlapping the end of the weld prep and on the top surface of the bottom inner forging
overlapping the outside diameter of the forging (adjacent to the outside diameter of the inner
shell). The backing bars prevent the lead contamination of the welds when the outer shell/bottom
outer forging weld and the bottom outer forging/bottom inner forging weld are performed after
cask body cooldown following the lead pour. Lead pouring preparations, the pour itself, and the
cooldown are performed in accordance with the lead pour requirements and procedures as
described in Section 8.4.2.
Following cooldown, the cask may be moved to a location that is more suitable for the
fabrication activities that are to follow. The temporary dam extension and supports at the open
8.4-2
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Docket No. 71-9235, Revision I 7C
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Revision l 8B
end of the outer shell are removed and the lead is machined to its final configuration, including
facing off the backing bars to ensure that no lead remains on the weld side of the backing bars.
The bottom outer forging is welded to the outer shell and to the bottom inner forging with the
backing bars preventing lead contamination of the welds. The weld examination and acceptance
criteria are described in Section 8.1.1 and defined on the License Drawings. The NS-4-FR
neutron shield material is installed in the bottom forging of the NAC-STC. The NS-4-FR is
machined to obtain the specified 2-inch thickness and to provide a groove around the outside
diameter. A backing bar is tack-welded on the inside diameter of the bottom outer forging in the
groove in the NS-4-FR and flush with its surface. The bottom plate is positioned and welded to
the bottom outer forging. The weld examination and acceptance criteria are described in Section
8.1.1 and are defined on the License Drawings.
The outside diameter of the outer shell is then machined to the specified final dimensions. If
required to achieve dimensional compliance with the License Drawings, additional localized
machining of the inner shell will be performed. Remachined areas of the inner shell shall be
re-examined by ultrasonic testing to confirm that the minimum thickness of 1.46 inches (37.1
mm) is maintained. Upon completion of final machining and prior to removal from the machine,
the dimensional inspection of the inside diameter and cylindricity of the cavity shall be
performed. Using calibrated M&TE, the inside diameter at 0, 45, 90 and 135 degree radial
locations shall be measured. This measurement shall be repeated at a minimum of 6 axial
locations through the bore of the inner shell. Using calibrated M&TE, a "sweep" of the entire
length of the bore at the same radial locations previously measured and also a "sweep" of the
diameter at the same axial locations will be performed. The combination of these two inspections
will demonstrate the actual diameter and cylindricity of the inner shell bore. Calibrated
inspection equipment and approved written procedures will be used to perform the final
dimensional inspections.
The Type 17-4 PH stainless steel lifting trunnions are welded to the top forging. The Type 17-4
PH stainless steel rotation trunnion recesses are welded to the outer shell at its juncture with the
bottom outer forging. Both the lifting trunnion and rotation trunnion recess weld surfaces are
prepared with a minimum 0.25-inch thick overlay of Inconel. The shear ring and the neutron
shield upper end plate are welded to the top forging. The weld examination and acceptance
criteria are described in Section 8.1.1, and are defined on the License Drawings .
8.4-3
NAC-STC SAR
Docket No. 71-9235
March 2017
Revision 18
The explosively-bonded stainless steel/copper (SS/Cu) heat transfer fins extending through the
neutron shield are welded ( only the stainless steel is welded) to the upper end plate and to the
outer shell. Following liquid penetrant examination of the fin to outer shell welds, the 24 neutron
shield shell plates are prepared for installation and 1/8-inch thick expansion foam is applied to
the interior surface using approved adhesive in accordance with the License Drawings. The
neutron shield shell plates are individually positioned and welded to the stainless steel extended
tip of the SS/Cu fins. These closure welds are then examined and accepted in accordance with
the requirements of the License Drawings. The cask is then placed in the inverted position
(closure end down). Following an installation procedure that has been approved by NAC or by
the material supplier if the material supplier is not NAC, the NS-4-FR neutron shield material is
installed by pouring into each of the 24 regions between the fins in the NAC-STC neutron shield
cavity. After the NS-4-FR has hardened, expansion foam (Section 4.5.3) is installed in the open
end of the neutron shield. The inside and outside diametrical ( curved) surfaces of the expansion
foam are covered by a protective thermal insulation material (Fiberfrax, see Section 4.5.4). The
24 sections of the neutron shield bottom end plate are each positioned and welded to the outer
shell, the fins, the neutron shield shell, and to each other. All of the neutron shield and fin welds
are liquid penetrant examined and accepted in accordance with the License Drawings. The
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neutron shield tank is leak tested using the pneumatic bubble method to verify shell integrity. •
The Type 17-4 PH stainless steel outer lid forging and the Type 304 stainless steel inner lid
forging are machined to the specified final dimensions. The NS-4-FR neutron shield material is
installed in the top of the inner lid following an installation procedure that has been approved by
NAC and by the material supplier. The exposed surface of the NS-4-FR is machined to obtain
the specified 2-inch thickness and the coverplate is welded to the inner lid body. The weld
examination and acceptance are in accordance with the requirements of the License Drawings.
The top surface of the inner lid is then final machined.
The remaining fabrication details (including the installation of the drain line) are then completed.
Following machining of the structural steel support disks and the aluminum heat transfer disks,
the components will be individually inspected for dimensional compliance to the License
Drawings to ensure that each disk meets the stated tolerances. The diameter of each disk is
measured using a calibrated external micrometer. The openings in each disk are inspected using
a calibrated three coordinate measurement machine. The machining center may also be used for
these inspections if previously qualified and calibrated. In the case of the diametral tolerances of
8.4-4 •
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NAC-STC SAR June 2018
Revision 18B Docket No. 71-9235, Revision 17B
the disks, the inspections are performed at 65 ± 5°F (18 ± 3°C) or else thermal expansion
corrections will be addressed during the inspection process.
The separately fabricated and assembled fuel basket is then inserted into the cask body by
carefully guiding the pre-assembled basket into the cask cavity. The acceptance tests described in
Section 8.1, not previously completed during fabrication, are performed and the completed
NAC-STC is prepared for delivery.
8.4.2 Description of Lead Pour Procedures (Standard Method)
This section describes the general requirements and the standard method procedure that applies
to the pouring of the lead in the annulus formed by the inner and outer shells of the NAC-STC
cask body. The lead annulus provides the primary radial gamma shielding in the cask body and is
subjected to a gamma scan test to verify its shielding integrity. The description that follows
includes the pre-pour preparations, the pouring of the molten lead in the annulus between the
inner and outer shells of the NAC-STC, and the post-pour controlled cooldown of the cask .
8.4.2.1 Preparation for Lead Pour
The following activities must be completed m preparation for pourmg of the lead m the
NAC-STC cask body:
1. Temporary stiffener bars/rings are installed both inside and outside of the body
weldment at intermittent locations along the cask length. The stiffeners support the
inner and outer shells during the lead pour and cooldown in order to maintain the
specified dimensions of the lead annulus. The stiffeners are removed after the
cooldown operation is completed.
2. A minimum of 12 pairs of thermocouples are used to monitor the heating and cooling
cycle of the inner and outer shells. Each pair of thermocouples is positioned at
approximately the same radial and axial location, one on the inside diameter of the
inner shell and one on the outside diameter of the outer shell.
3. Electric heaters are installed in the cask cavity for use in heating the inner shell.
8.4-5
NAC-STC SAR
Docket No. 71-9235, Revision l 7B
June 2018
Revision l 8B
4. The body weldment (Section 8.4.1) of the NAC-STC is inverted and suppo1ied in a
stable, vertical position in a "pit" or within a windbreak structure to provide a
basically draft-free operations area.
5. An auxiliary dam extension and supports are welded to the open end of the outer
shell. The extension and supports permit the full length of the lead shell to be poured
in one operation while maintaining the annulus spacing at the open end of the outer
shell.
6. A minimum of 20 gas heating/water cooling rings are installed around the outside of
the body weldment for use in heating, and later in cooling, the outer shell. Gas torches
are provided for heating the outside surface of the bottom inner forging.
7. The body weldment surfaces, especially the lead annulus, are checked for dimensional
accuracy to ensure that the required spacing has been maintained and for cleanliness
to ensure that no foreign materials are present.
•
8. The typical general arrangement of the equipment for the standard lead pour •
operation is shown in Figure 8.4-1.
8.4.2.2 Lead Pour Operations
The requirements and activities that must be completed during the pouring of the lead in the
NAC-STC cask body are:
1. The lead material certification is checked to ensure that it conforms to the
requirements of the American Society of Testing Materials (ASTM) B29, Chemical
Copper Grade - 99.90 percent pure.
2. Approximately 60,000 pounds of lead is placed in appropriate size kettles and melted.
During the lead pouring operations the temperature of the molten lead is maintained
between 650°F (343°C) and 750°F (399°C).
3. At the same time that the lead is being melted, the NAC-STC body weldment is
simultaneously heated using both the electric heaters on the interior and the gas
8.4-6 •
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NAC-STC SAR June 2018
Revision 18B Docket No. 71-9235, Revision l 7B
8.4.2.3
heating rings on the exterior. The body weldment will be heated in a steady and
uniform manner at a rate not exceeding 125°F/hour (69.4°C/hour). Gas torches are
used to heat the exterior of the bottom inner forging. The surface temperature of the
body weldment is never permitted to exceed 800°F ( 427°C). The temperature of the
entire body weldment is maintained between 640°F (338°C) and 740°F (393°C)
throughout the lead pour operations.
4. The lead pour is initiated immediately after the temperatures of the lead and the body
weldment are stabilized in the ranges previously specified. The actual pouring of the
lead is completed without interruption and in as short a period of time as possible.
During the lead pour the bottom end of the filler-tube is kept below the surface of the
molten lead to preclude the formation of voids in the lead.
5. The lead is poured to a level that is sufficient to ensure that dross removal and
contraction during solidification do not reduce the finished surface below the required
level. A long steel rod inserted into the molten lead annulus is used to ensure that no
solidification has begun anywhere in the volume of molten lead .
Cooldown Following Lead Pour
The procedures and requirements that must be completed during cooldown of the NAC-STC
body weldment following completion of the lead pour are as follows:
1. Cooldown is initiated by turning off the electrical heater (interior) and the gas
heating/water cooling ring (exterior) at the lowest end of the cask (in the as-poured
position). The gas heating/water cooling ring is then used to facilitate and control
cooling by spraying water on the exterior surface of the cask. As cool down proceeds,
the heaters and rings upward along the cask are successively turned off and the
cooling water spray is turned on from each ring.
2. The cooldown process is temperature controlled to maintain approximately uniform
solidification conditions across the thickness and around the circumference of the
annulus .
8.4-7
NAC-STC SAR June 2018
Revision l 8B Docket No. 71-9235, Revision l 7B
8.4.2.4
3. The cooldown rate is held steady and uniform at a rate not to exceed 125°F/hour
(69.4°C/hour) and the temperature differential between the inside shell and the
outside shell is not allowed to exceed 100°F (55.5°C). Once the inner and outer shell
temperatures have cooled to l 50°F (66°C), it is no longer necessary to control the
cooldown rate.
4. The solidification level in the lead annulus is checked with the aid of a long steel rod.
The maximum difference in the elevation of the solidified lead between the inside
surface of the outer shell and the outside surface of the inner shell is not permitted to
exceed 2 inches (51 mm).
5. Dross is skimmed off the top of the lead while maintaining the molten head
throughout the cooldown process.
Lead Pour Documentation
The following data is included in the Data Package for the Lead Pour Operation:
1. Certificate of Chemical Analysis of the lead.
2. Heating and cooling charts showing elapsed time and temperatures.
3. Location, time and temperature for readings taken with a handheld pyrometer or other
temperature reading device.
4. Difference in solidification elevations when checking at the inside surface of the outer
shell and the outside surface of the inner shell.
8.4.3 Description of Lead Pour Procedures (Alternate Method)
This section describes the general requirements and the alternate method procedure that applies
to the pouring of the lead in the annulus formed by the inner and outer shells of the NAC-STC
cask body. The lead annulus provides the primary radial gamma shielding in the cask body and is
subjected to a gamma scan test to verify its shielding integrity. The description that follows
includes the pre-pour preparations, the pouring of the molten lead in the annulus between the
inner and outer shells of the NAC-STC, and the post-pour controlled cooldown of the cask .
8.4-8
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NAC-STC SAR June 2018
Revision 18B Docket No. 71-9235, Revision 17B
8.4.3.1 Preparation for Lead Pour
The following activities must be completed 111 preparation for pour111g of the lead 111 the
NAC-STC cask body:
1. Temporary stiffener bars/rings are installed inside the body weldment at intermittent
locations along the cask length. Optional stiffener bars/rings may be installed on the
outside of the body weldment at intermittent locations along the cask length. The
stiffeners support the shells during the lead pour and cooldown in order to maintain
the specified dimensions of the lead annulus. The stiffeners are removed after the
cooldown operation is completed.
2. Pairs of thermocouples are used to monitor the heating and cooling cycle of the inner
and outer shells. Each pair of thermocouples is positioned at approximately the same
radial and axial location, one on the inside diameter of the inner shell and one on the
outside diameter of the outer shell. The exact number of pairs shall be determined
prior to the lead pour .
3. Heaters are installed in the cask cavity for use in heating the inner shell. Typical
heaters include but are not limited to electric, gas, etc.
4. The body weldment (Section 8.4.1) of the NAC-STC is inverted and supported in a
stable, vertical position within a structure that provides a basically draft-free
operational area.
5. An auxi I iary dam extension and supports are welded to the open end of the outer
shell. The extension and supports permit the full length of the lead shell to be poured
in one operation while maintaining the annulus spacing at the open end of the outer
shell.
6. Heating/water cooling systems are installed around the outside of the body weldment
for use in heating, and later in cooling, the outer shell. A heating system is also used
on the outside surface of the bottom inner forging. Heating methods include but are
not limited to electric, gas, etc. Cooling methods include but are not limited to
cooling rings, cooling shells, etc. encompassing the inner and outer shells.
7. The body weldment surfaces, especially the lead annulus, are checked for
dimensional accuracy to ensure that the required spacing has been maintained and for
cleanliness to ensure that no foreign materials are present.
8.4-9
NAC-STC SAR June2018
Revision l 8B Docket No. 71-9235, Revisions l7B and 18A
8.4.3.2
8. The typical general arrangement of the equipment for the alternate method lead pour
operation is shown in Figure 8.4-2.
Lead Pour Operations
The requirements and activities that must be completed during the pouring of the lead in the
NAC-STC cask body are:
1. The lead material certification is checked to ensure that it conforms to the
requirements of the American Society of Testing Materials (ASTM) B29, Chemical
Copper Grade - 99.90 percent pure.
2. Approximately 60,000 pounds of lead is placed in appropriate size kettles and melted.
During the lead pouring operations the temperature of the molten lead is maintained
above a sufficient temperature to conduct the pour but below 790°F ( 421 °C).
3. At the same time that the lead is being melted, 1 the NAC-STC body weldment is
simultaneously heated on both the interior and exterior. The body weldment will be
heated in a steady, uniform, and controlled manner at a rate not exceeding 125°F/hour
(69.4°C/hour). A heating system is also used to heat the exterior of the bottom inner
forging. The surface temperature of the body weldment is never permitted to exceed
800°F (427°C). The temperature of the entire body weldment is maintained between
640°F (338°C) and 740°F (393°C) throughout the lead pour operations.
4. The lead pour is initiated after the temperatures of the lead and the body weldment
are stabilized, as previously described. The actual pouring of the lead is completed
without interruption and in as short a period of time as possible. During the lead pour
the bottom end of the filler-tube is kept below the surface of the molten lead to
preclude the formation of voids in the lead.
5. The lead is poured to a level that is sufficient to ensure that dross removal and
contraction during solidification do not reduce the finished surface below the required
level. A long steel rod inserted into the molten lead annulus is used to ensure that no
solidification has begun anywhere in the volume of molten lead.
8.4-10
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NAC-STC SAR June2018
Revision l 8B Docket No. 71-9235, Revisions l 7B and 18A
8.4.3.3 Cooldown Following Lead Pour
The procedures and requirements that must be completed during cooldown of the NAC-STC
body weldment following completion of the lead pour are as follows:
I. Cooldown is initiated by turning off the interior and exterior heaters at the lowest end
of the cask (in the as-poured position). A water tank cooling system is then used to
facilitate and control cooling. As cask cooling proceeds, effected by raising the
cooling system water level, the heaters upward along the cask are successively turned
off, ensuring the head of lead during cooling remains in the molten state.
2. The cooldown process is temperature controlled to maintain approximately uniform
solidification conditions across the thickness and around the circumference of the
annulus. This is maintained by controlling both the minimum inlet water temperature
and the rate in which the water level is increased. The maximum water level rate of
increase shall be no greater than 23.62 inches (600mm)/hr and the minimum inlet
water temperature shall be greater than 45°F (7°C). Overall cooling rate(s) should be
controlled by lead solidification measurements described below in Step 3 below.
3. The solidification level in the lead annulus is checked with the aid of a long steel rod
in order to verify there is no significant difference in solid surface between the inside
and outside of the annulus.
4. Dross is skimmed off the top of the lead while maintaining the molten head
throughout the cooldown process.
8.4.3.4 Lead Pour Documentation
The following data is included in the Data Package for the Lead Pour Operation:
I. Certificate of Chemical Analysis of the lead.
2. Heating and cooling charts showing elapsed time and temperatures .
8.4-11
NAC-STC SAR
Docket No. 71-9235, Revision l 7B
Figure 8.4-1 Typical Arrangement of Lead Pour Equipment (Standard Method)
SCAFFOLD
0 0 0
0 0 0 0
ELECTRICAL HEATERS
AUXIWARY DAM
8.4-12
0 0
HEATING &: COOLING RINGS
STIFFENER RINGS
June 2018
Revision l 8B •
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NAC-STC SAR
Docket No. 71-9235, Revision 17B
Figure 8.4-2 Typical Arrangement of Lead Pour Equipment (Alternate Method)
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lHEFiMOCUPLES ,
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,I ,!
i ·1
8.4-13
June 2018
Revision 18B
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