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{{#Wiki_filter:Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 4.0 CONTAINMENT 4.1 Description of the Containment System The containment function of the ATR FFSC is to confine the fuel elements or loose plates within the packaging during Normal Conditions of Transport (NCT) and Hypothetical Accident Conditions (HAC).
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The package body is a stainless steel weldment that consists of two nested shells. The outer shell is an 8-in square stainless steel tube with a 3/16-in thick wall, and the inner shell is a 6-in diameter stainless steel tube with a 0.120-in thick wall. Components are joined using full-thickness fillet welds (i.e., fillet welds whose leg size is nominally equal to the lesser thickness of the parts joined) and full and partial penetration groove welds. The end of the body is welded closed with 0.88-in plate.
The lid end of the package is closed with a simple closure device. The closure engages with the body using a bayonet style design. There are four lugs, uniformly spaced on the closure, that engage with four slots in the mating body fixture. The closure is secured by two retracting spring loaded pins, rotating the closure through 45°, and releasing the spring loaded pins such that the pins engage with the mating holes on the body. When the pins are properly engaged with the mating holes the closure is locked and cannot be removed unintentionally.
The containment boundary is defined as the boundary of the cavity formed by the closure and inner stainless steel tube. For criticality control purposes, the fuel element must remain within this boundary during NCT and HAC. No seals or gaskets are utilized within the package.
To prevent unauthorized operation, a small post on the closure is drilled to receive a tamper indicating device (TID) wire. An identical post is located on the body and is also drilled for the TID wire. For ease in operation, there are two TID posts on the body. There are only two possible angular orientations for the closure installation and the duplicate TID post on the body enables TID installation in both positions.
4.1.1 Type A Fissile Packages The ATR FFSC is classified as a Type A Fissile package. The Type A Fissile package is constructed and prepared for shipment so that there is no loss or dispersal of the radioactive contents, and no significant increase in external surface radiation levels, and no substantial reduction in the effectiveness of the packaging during normal conditions of transport. The fissile material is contained within the containment boundary. Chapter 6.0, Criticality Evaluation, demonstrates that the package remains subcritical under normal and hypothetical accident conditions.
The ATR FFSC contains four radioactive isotopes: U-234, U-235, U-236, and U-238. The A2 value for U-235 and U-238 is unlimited, while the minimum A2 value for U-234 and U-236 is 0.16 Ci for slow lung absorption. To compute the mixture A2 for the HEU payloads, the maximum value of 1200 g U-235 is assumed, with a low weight fraction of 90% to maximize the mass of uranium. Therefore, the total mass of uranium is 1200/0.9 = 1333 g U. The maximum weight percents of U-234 (1.2%) and U-236 (0.7%) are assumed to maximize the mass of these isotopes. The balance is treated as U-238. For this conservative isotopic mix, the mixture A2 is 4-1
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                Rev. 16, May 2021 0.164 Ci. The package activity for this mixture is 0.103 Ci (mostly due to U-234); therefore, the package contains approximately 0.6A2.
4.1.2 Type B Packages The content of the ATR FFSC package is high-enriched uranium with approximately 0.6A2 for release purposes. As a fissile package the ATR FFSC must meet the release rates for Type B packages when required by the total amount of radioactive material. However, because the A2 value of the contents is less than 1 A2, the package is classified as Type A and there are no release limits except as necessary for criticality control.
4.2 Containment under Normal Conditions of Transport The ATR FFSC payloads listed in Section 1.2.2, Contents, are confined within the packaging under NCT. This is verified by full-scale testing, as discussed in Section 2.6, Normal Conditions of Transport. The test units survived the NCT drop tests with minimal damage to the packaging and no damage to the fuel elements. The maximum internal pressure in the package does not exceed atmospheric pressure because the closure is not sealed with a gasket or other sealing material. Because the ATR FFSC is a Type A Fissile package, leakage rate testing is not required.
4.3 Containment under Hypothetical Accident Conditions The radioactive material contents of the ATR FFSC package must meet the containment requirements of 10 CFR §71.55(e) such that the package would be subcritical under the HAC.
The test program demonstrates that the package contains the fuel elements or loose fuel plates under the HAC events sufficient to maintain criticality control. The full-scale HAC drop tests summarized in Section 2.7, Hypothetical Accident Conditions, confirm the HAC performance of the package. The closure remained intact throughout all the drop sequences, and the fuel element remained confined within the inner stainless steel tube. The non-fissile end boxes on the fuel element shattered as expected but the fueled portion of the element remained intact and retained its geometry. There was no dispersal of fissile material. The criticality evaluation presented in Section 6.0, Criticality Evaluation, evaluates the contents in the most reactive credible configuration and with water moderation as required.
Because the ATR FFSC package is a Type A Fissile package and the contents are less than 1 A2, the performance requirements of 10 CFR §71.51 do not apply.
4.4 Leakage Rate Tests for Type B Packages The ATR FFSC is a Type A Fissile package; therefore, this section does not apply.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                          Rev. 16, May 2021 5.0 SHIELDING EVALUATION Compliance of the ATR FFSC with respect to the dose rate limits established by 10 CFR §71.471 for normal conditions of transport (NCT) or 10 CFR §71.51(a)(2) for hypothetical accident conditions (HAC) are satisfied when limiting the package to the contents specified in Section 1.2.2, Contents, and verified by measurement.
Prior to transport, the ATR FFSC shall be monitored for both gamma and neutron radiation to demonstrate compliance with 10 CFR §71.47. Although the ATR FFSC will likely be shipped exclusive use, dose rates will be sufficiently low to allow non-exclusive use transport, if desired.
Shielding materials are not specifically provided by the ATR FFSC. Because the contents are essentially unshielded, the HAC dose rates at one meter will not be significantly different from the NCT dose rates at one meter. This result ensures that the post-HAC, allowable dose rate of 1 rem/hr a distance of one meter from the package surface per 10 CFR §71.51(a)(2) will be met.
1 Title 10, Code of Federal Regu1ations, Part 71 (10 CFR 71), Packaging and Transportation of Radioactive Material.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                            Rev. 16, May 2021
 
==6.0 CRITICALITY EVALUATION==
 
The following analyses demonstrate that the ATR FFSC complies with the requirements of 10 CFR §71.551 and §71.59.
The analysis in the main body of Chapter 6 (Sections 6.2 through 6.6, 6.8, and 6.9) pertains only to the ATR high enriched uranium (HEU) fuel element and ATR loose plate basket. Any discussion of the ATR fuel element or ATR FHE in these sections is limited solely to the ATR HEU fuel element and ATR HEU FHE. Additional payloads are added as appendices.
The analysis for MIT and MURR HEU fuel is contained in Appendix B (Section 6.10, Criticality Analysis for MIT and MURR Fuel). Any discussion of the MIT and MURR fuel elements and MURR FHE in these sections is limited solely to the MIT and MURR HEU fuel elements and the MURR HEU FHE.
The analysis for the small quantity payloads is contained in Appendix C (Section 6.11, Criticality Analysis for Small Quantity Payloads).
The analysis for the ATR U-Mo demonstration element was previously contained in Appendix D (Section 6.12, Criticality Analysis for the U-Mo Demonstration Element). This analysis and any associated results have been removed as the ATR U-Mo demonstration element is no longer a possible payload.
The analysis for the Cobra fuel element is contained in Appendix E (Section 6.13, Criticality Analysis for the Cobra Fuel Element).
The analysis for the ATR low enriched uranium (LEU) fuel element, MURR LEU fuel element and design demonstration element (DDE), MIT LEU fuel element and DDE, and NBSR DDE is contained in Appendix F (Section 6.14, Criticality Analysis for ATR, MURR, MIT, and NBSR LEU Fuel Elements and/or DDEs).
The air transport analysis in Section 6.7 applies to all payloads except those documented in Section 6.14 (for which the applicable air transport analysis is also documented in Section 6.14).
6.1 Description of Criticality Design The results presented in this section are for all payload types.
6.1.1 Design Features Important for Criticality A comprehensive description of the ATR FFSC is provided in Section 1.2, Packaging Description, and in the drawings in Appendix 1.3.2, Packaging General Arrangement Drawings.
This section summarizes those design features important for criticality.
No poisons are utilized in the package.
For the ATR HEU fuel element payload, the separation provided by the packaging (outer tube minimum flat-to-flat dimension of 7.9-in, inner tube maximum inner diameter of 5.814-in), along with the limit on the number of packages per shipment, is sufficient to maintain criticality safety.
1 Title 10, Code of Federal Regulations, Part 71 (10 CFR 71), Packaging and Transportation of Radioactive Material.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                        Rev. 16, May 2021 For the ATR loose plate payload, in addition to the packaging design features noted above, moderation of the loose plates is controlled by the loose plate basket, which confines the fuel plates to a rectangular area.
For the MURR/MIT HEU fuel element payloads, in addition to the packaging design features noted above, the MURR HEU and MIT fuel handling enclosures (FHEs) restrict postulated fuel element pitch expansion under hypothetical accident conditions.
For the small quantity payload analysis, the fuel is modeled as a homogenized mixture of uranium and water. Therefore, the packaging itself is sufficient to maintain criticality safety, as the fuel handling enclosures and fuel structural materials are not credited in the analysis.
For the Cobra fuel element analysis, in addition to the packaging design features noted above, the Cobra FHE is credited for normal conditions of transport.
For the MURR/MIT LEU fuel element and DDE payloads, in addition to the packaging design features noted above, the MURR LEU FHE and MIT FHE restrict postulated fuel element pitch expansion and movement under hypothetical accident conditions. For the ATR LEU fuel element and NBSR DDE payloads, the applicable fuel handling enclosure or protective material is conservatively not credited.
6.1.2 Summary Table of Criticality Evaluation The upper subcritical limit (USL) for ensuring that the ATR FFSC (single package or package array) is acceptably subcritical, as determined in Section 6.8 (for HEU plate-fuel), Section 6.11.8 (for the small quantity payload), and Section 6.13.8 (for the Cobra fuel element) is:
USL = 0.9209 The USL for ATR, MURR, MIT, and NBSR LEU fuel elements and/or DDEs is calculated on a case-by-case basis in Section 6.14.6.
The package is considered to be acceptably subcritical if the computed ksafe (ks), which is defined as keffective (keff) plus twice the statistical uncertainty (), is less than or equal to the USL, or:
ks = keff + 2  USL The USL is determined on the basis of a benchmark analysis and incorporates the combined effects of code computational bias, the uncertainty in the bias based on both benchmark-model and computational uncertainties, and an administrative margin. The results of the benchmark analysis indicate that the USL is adequate to ensure subcriticality of the package.
ATR HEU Fuel Element and ATR Loose Plate Basket The packaging design is shown to meet the requirements of 10 CFR 71.55(b) when the package is limited to either one 1200 g U-235 ATR HEU fuel element, or 600 g U-235 in the form of ATR loose fuel plates. Moderation by water in the most reactive credible extent is utilized in both the NCT and HAC analyses. In the single package NCT models, full-density water fills the accessible cavity, while in the single package HAC models, full-density water fills all cavities.
In the fuel element models, the most reactive credible configuration is utilized by maximizing the gap between the fuel plates. Maximizing this gap maximizes the moderation and hence the reactivity because the system is under moderated. In the loose plate model, no credit is taken for 6-2
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 the dunnage plates and the optimal pitch and fuel arrangement is utilized. In all single package models, 12-in of water reflection is utilized.
In the NCT and HAC array cases, partial moderation is considered to maximize array interaction effects. A 9x9x1 array is utilized for the NCT array, while a 5x5x1 array is utilized in the HAC array. In all array models, 12-in of water reflection is utilized.
The maximum results of the ATR HEU fuel element criticality calculations are summarized in Table 6.1-1. The maximum calculated ks is 0.8362, which occurs for the optimally moderated NCT array case. The NCT array is more reactive than the HAC array because the NCT array is larger, and moderation is allowed in both conditions. In this case, the fuel element is moderated with full-density water, the inner tube is moderated with 0.3 g/cm3 water, and void is modeled between the insulation and outer tube.
The maximum results of the loose plate basket criticality calculations are summarized in Table 6.1-2. The maximum calculated ks is 0.7747, which occurs for the optimally moderated NCT array case. The NCT array is more reactive than the HAC array because the NCT array is larger, and moderation is allowed in both conditions. In this case, the loose fuel plate basket is moderated with full-density water, the inner tube is moderated with 0.5 g/cm3 water, and void is modeled between the insulation and outer tube.
It may be noted when comparing Table 6.1-1 and Table 6.1-2 the fuel element payload is more reactive than the loose plate basket payload.
MURR and MIT HEU Fuel Element A summary of the MURR and MIT HEU fuel element analysis is provided in Section 6.10.1.2.
The summary table is also provided as Table 6.1-3 for convenience.
Small Quantity Payload A summary of the small quantity payload analysis is provided in Section 6.11.1.2. The summary table is also provided as Table 6.1-4 for convenience.
Cobra Element A summary of the Cobra element analysis is provided in Section 6.13.1.2. The summary table is also provided as Table 6.1-5 for convenience.
Air Transport Analysis The air transport analysis applies to all licensed payloads. In the air transport analysis, 2000 g U-235 is modeled as a sphere moderated with the hydrogenous packaging materials and reflected with 20 cm of water. The hydrogenous packaging materials include 100 g polyethylene and 4000 g neoprene. Maximum reactivity is achieved when the fissile material is divided into a 1500 g U-235 inner sphere moderated with polyethylene and neoprene and an outer sphere consisting of 500 g U-235 uranium metal. The maximum calculated ks is 0.6074 for the most reactive air transport case, which is far below the USL.
ATR, MURR, MIT, and NBSR LEU Fuel Elements and/or DDEs A summary of the ATR, MURR, MIT, and NBSR LEU fuel elements and/or DDEs is provided in Section 6.14.1.2. The summary table is also provided as Table 6.1-6 for convenience.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                  Rev. 16, May 2021 Table 6.1 Summary of Criticality Evaluation (ATR Fuel Element Payload)
Normal Conditions of Transport (NCT)
Case                      ks Single Unit Maximum            0.4224 9x9 Array Maximum              0.8362 Hypothetical Accident Conditions (HAC)
Case                      ks Single Unit Maximum            0.4524 5x5 Array Maximum              0.7453 USL = 0.9209 Table 6.1 Summary of Criticality Evaluation (ATR Loose Plate Payload)
Normal Conditions of Transport (NCT)
Case                      ks Single Unit Maximum            0.4020 9x9 Array Maximum              0.7747 Hypothetical Accident Conditions (HAC)
Case                      ks Single Unit Maximum            0.4363 5x5 Array Maximum              0.6979 USL = 0.9209 Table 6.1 Summary of Criticality Evaluation (MURR/MIT Payload)
MURR            MIT Normal Conditions of Transport (NCT)
Case                        ks            ks Single Unit Maximum              0.44807        0.36978 9x9 Array Maximum                0.85643        0.65658 Hypothetical Accident Conditions (HAC)
Case                        ks            ks Single Unit Maximum              0.54584        0.43666 5x5 Array Maximum                0.85881        0.67309 USL = 0.9209 6-4
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                  Rev. 16, May 2021 Table 6.1 Summary of Criticality Evaluation (Small Quantity Payload)
Normal Conditions of Transport (NCT)
Case                  ks Single Unit Maximum          0.6478 10 Package Array Maximum        0.8943 Hypothetical Accident Conditions (HAC)
Case                  ks Single Unit Maximum          0.7244 4 Package Array Maximum          0.8222 USL = 0.9209 Table 6.1 Summary of Criticality Evaluation (Cobra Element)
Normal Conditions of Transport (NCT)
Case                      ks 71.55(b)(d): Single Unit Maximum            0.4032 71.55(d), 71.59: 9x9x1 Array Maximum        0.7389 Hypothetical Accident Conditions (HAC)
Case                      ks 71.55(e): Single Unit Maximum                0.4996 71.55(e), 71.59: 5x5x1 Array Maximum        0.7643 USL = 0.9209 6-5
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                              Rev. 16, May 2021 Table 6.1 Summary of Criticality Evaluation (LEU Fuel Elements and DDEs)
ATR LEU        MURR LEU      MIT LEU      NBSR LEU Normal Conditions of Transport (NCT)
Case                  ks                ks          ks              ks Single Unit Maximum        0.42385            0.5063      0.42693        0.39137 0.77809          0.86146      0.68613        0.65257 Array Maximum (7x7)              (8x8)      (9x9)          (9x9)
Hypothetical Accident Conditions (HAC)
Case                  ks                ks          ks              ks Single Unit Maximum        0.58853          0.58572      0.52045        0.50968 0.87986          0.86625      0.72735        0.74897 Array Maximum (4x4)              (5x5)      (5x5)          (5x5)
Upper Subcritical Limit (USL) 0.92312          0.92071      0.92288        0.92348 Air Transport Analysis ks = 0.71171                USL = 0.90868 6.1.3 Criticality Safety Index The HAC array calculations are performed for 2N packages and the NCT array calculations are performed for at least 5N packages. The number of packages modeled for each payload type is summarized in Table 6.1-7, along with the value of N. Note that for many of the NCT array cases, the number of packages modeled conservatively exceeds the 5N value. The 10 CFR
§71.59 criticality safety index (CSI) is computed as 50/N and is provided in Table 6.1-7 for each payload type.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                  Rev. 16, May 2021 Table 6.1 Summary of Criticality Safety Indices NCT Array      HAC Array Payload                    (# Packages)    (# Packages)      N          CSI ATR HEU fuel element ATR loose plates MIT HEU fuel element            81            25          12.5          4.0 MURR HEU fuel element Cobra fuel element Small Quantity Payload          10              4          2.0          25.0 ATR LEU fuel element            49            16          8.0          6.25 MURR LEU fuel element and DDE MIT LEU fuel element            64            25          12.5          4.0 and DDE NBSR DDE 6-7
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 6.2 Fissile Material Contents The package can accommodate either (i) one ATR Mark VII fuel element, or (ii) a loose plate basket filled with ATR Mark VII fuel plates.
6.2.1 Fuel Element Four different ATR Mark VII fuel element types are available: standard (7F), non-borated (7NB), non-borated hybrid (7NBH), and non-fueled plate 19 (YA). These fuel element types are described in Section 1.2.2, Contents. The 7NB fuel element is the only fuel element that does not contain boron, and is conservatively utilized in the criticality analysis.
Each fuel element contains up to 1200 g U-235, enriched up to 94 wt.%. The U-235 mass per plate is provided in Table 6.2-1. These values are generated by scaling up the U-235 loading for a 1075 g U-235 fuel element, as the 1200 g limit has been selected to envelope future increases in the loading. The weight percents of the remaining uranium isotopes are 1.2 wt.% U-234 (max), 0.7 wt.% U-236 (max), and 5.0-7.0 wt.% U-238. Each fuel element contains 19 curved fuel plates. Fuel plate 1 has the smallest radius, while fuel plate 19 has the largest radius, as shown in Figure 6.2-1. The as-modeled fuel element is shown in Figure 6.2-2. The fuel meat is uranium aluminide (UAlx) mixed with additional aluminum. In the following paragraphs, the details of the fuel element are provided.
The key fuel element dimensions and tolerances utilized in the criticality models are summarized on Figure 6.2-1. Fuel plate 1 is nominally 0.080-in thick, fuel plates 2 through 18 are nominally 0.050-in thick, and fuel plate 19 is nominally 0.100-in thick. The plate thickness tolerance is
+0.000/-0.002-in for all plates. The fuel meat is nominally 0.02-in thick for all 19 plates. The plate cladding material is aluminum ASTM B 209, 6061-0. Fuel element side plates are fabricated of ASTM B 209, aluminum alloy 6061-T6 or 6061-T651. All aluminum alloys are modeled as pure aluminum. The fuel element side plates have a minimum thickness of 0.182-in.
Channels 2 through 10 have a width of 0.078 +/- 0.007-in, while channels 11 through 19 have a width of 0.077 +0.008/-0.006-in. These tolerances represent average and not localized channel width. Therefore, the maximum average channel width is 0.085-in. For an actual fuel element, the channel width may exceed these tolerances in localized areas. The local maximum is 0.087-in.
The arc length of the fuel meat changes from plate to plate. This arc length varies based on the distance from the edge of the fuel meat to the fuel element side plate, as defined for each plate on Figure 6.2-1. This dimension is 0.245-in (max)/0.145-in (min) for fuel plates 1 and 19, 0.145-in (max)/0.045-in (min) for fuel plates 2 through 17, and 0.165-in (max)/0.065-in (min) for fuel plate 18. The smaller this dimension, the larger the arc length of the fuel meat.
The active fuel length varies between a minimum of 47.245-in (= 49.485 - 2*1.12) and a maximum of 48.775-in (= 49.515 - 2*0.37) for all fuel plates.
It is demonstrated in Section 6.4.1.2.1, Fuel Element Payload Parametric Evaluation, that reactivity increases with increasing meat arc length. Therefore, the arc length is modeled at the maximum value. To determine the number densities of the fuel meat, it is first necessary to compute the volume of the fuel meat. The volume of the fuel meat for each plate is the maximum arc length of the meat multiplied by the fuel length (48-in) and meat thickness (0.02-6-8
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                    Rev. 16, May 2021 in). The fuel length and meat thickness are treated as fixed quantities in all fuel element models, and the use of these dimensions is justified in Section 6.4.1.2.1.
The fuel meat volume for each of the 19 fuel plates is provided in Table 6.2-1. The mass of U-235 per plate utilized in the analysis is also provided in Table 6.2-1. The U-235 gram density for each fuel plate is also computed. Note that the U-235 gram density is higher in the inner plates compared to the outer plates.
The fuel itself is a mixture of UAlx and aluminum. The density of this mixture is proportional to the U-235 gram density, as shown in Table 6.2-2. These data are perfectly linear, and a linear fit of the data is 2 = 0.87331 + 2.5357, where 2 is the total gram density of the mixture, and 1 is the gram density of the U-235 in the mixture. This equation is used to compute the total mixture gram density provided as the last column in Table 6.2-1.
From the fuel volumes, U-235 gram densities, and total mixture densities provided in Table 6.2-1, the number densities for the fuel region of each fuel plate may be computed. These number densities are provided in Table 6.2-3. The U-235 weight percent is assumed to be the maximum value of 94%. Representative weight percents of 0.6% and 0.35% are assumed for U-234 and U-236, respectively, and the balance (5.05%) is modeled as U-238.
6.2.2 Loose Fuel Plates The loose plate basket may transport up to 600 g of U-235 in the form of ATR Mark VII fuel plates. These plates are described in Section 6.2.1, Fuel Element, although the loose plates may be flat as well as curved. The widths of the fuel meat for flat plates are the same as the fuel meat arc lengths provided in Table 6.2-1.
Because an integer number of plates will be transported, for computational purposes it is useful to modify the mass of U-235 per plate so that the total U-235 mass per package adds to 600 g.
The column labeled Number of Plates to 600 g in Table 6.2-4 is simply the total desired mass (600 g) divided by the mass of U-235 per plate (from Table 6.2-1) and gives an estimate of the number of plates of each type required to reach 600 g U-235. Detailed models are developed for only four plates: 3, 5, 8, and 15. It is demonstrated in the analysis that it is sufficient to bound all of the plates by modeling these four. The number of plates modeled and the modeled mass of U-235 per plate are provided as the last two columns in Table 6.2-4.
In fuel element calculations, it has been determined that the fuel element is the most reactive when the arc length of the fuel meat is maximized. Therefore, all loose plate models utilize fuel plates with maximized fuel meat arc length. Also, because it has been determined that nominal fuel meat thickness (0.02-in) and nominal active fuel length (48.0-in) may be utilized with negligible effect on the reactivity, all loose plate models utilize these nominal dimensions.
The overall plate thickness tolerance is +0.000/-0.002-in, and the loose plates are modeled at the minimum thickness of 0.048-in by reducing the cladding thickness by 0.001-in.
The number densities utilized in the models are provided in Table 6.2-5. These number densities are computed using the same method utilized in the fuel element models, although the U-235 mass per plate has been slightly adjusted as necessary so that the models always have 600 g U-235.
The active fuel length is modeled as 48-in for all fuel plates, consistent with the treatment of the fuel elements. The axial regions outside the active fuel region are conservatively ignored. The 6-9
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                Rev. 16, May 2021 width of cladding from the fuel meat to the edge of the plate is modeled as 0.045-in for all of the fuel plates, which is the minimum dimension from the fuel meat to the fuel element support structure. The actual plates are wider than modeled because the plates extend into the fuel element support structure and this additional width is neglected.
Table 6.2 Fuel Element Volume and Gram Densities Fuel Meat    Fuel Meat    U-235 Mass        U-235      Total UAlx +
Arc Length      Volume      Per Plate      density, 1  Al Density, 2 Plate        (cm)        (cm3)          (g)          (g/cm3)        (g/cm3) 1        4.2247        26.2          27.1            1.04          3.44 2        5.0209        31.1          32.5            1.04          3.45 3        5.2764        32.7          43.2            1.32          3.69 4        5.5319        34.3          45.1            1.32          3.69 5        5.7873        35.8          58.2            1.62          3.95 6        6.0427        37.4          60.9            1.63          3.96 7        6.2982        39.0          63.6            1.63          3.96 8        6.5536        40.6          66.3            1.63          3.96 9        6.8090        42.2          69.0            1.64          3.96 10        7.0644        43.8          71.7            1.64          3.97 11        7.3198        45.3          74.3            1.64          3.97 12        7.5752        46.9          77.0            1.64          3.97 13        7.8306        48.5          79.7            1.64          3.97 14        8.0860        50.1          82.4            1.64          3.97 15        8.3414        51.7          85.2            1.65          3.98 16        8.5968        53.2          71.4            1.34          3.71 17        8.8521        54.8          73.6            1.34          3.71 18        9.0058        55.8          60.1            1.08          3.48 19        8.9039        55.1          58.7            1.06          3.47 Total          --        824.5        1200.0            --            --
Table 6.2 Fuel Density Equation U-235 Density (g/cm3)          Total Fuel Density (g/cm3) 1                              2 1.00                            3.409 1.30                            3.671 1.60                            3.933 Linear Fit: 2 = 0.87331 + 2.5357 6-10
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                Rev. 16, May 2021 Table 6.2 Fuel Element Number Densities U-234          U-235            U-236          U-238          Aluminum          Total Plate (atom/b-cm)  (atom/b-cm)      (atom/b-cm)    (atom/b-cm)      (atom/b-cm)    (atom/b-cm) 1    1.7026E-05    2.6560E-03      9.8475E-06    1.4089E-04      5.2187E-02    5.5010E-02 2    1.7156E-05    2.6763E-03      9.9226E-06    1.4196E-04      5.2153E-02    5.4998E-02 3    2.1711E-05    3.3869E-03      1.2557E-05    1.7966E-04      5.0974E-02    5.4574E-02 4    2.1618E-05    3.3724E-03      1.2503E-05    1.7889E-04      5.0998E-02    5.4583E-02 5    2.6648E-05    4.1571E-03      1.5413E-05    2.2051E-04      4.9696E-02    5.4115E-02 6    2.6746E-05    4.1724E-03      1.5470E-05    2.2132E-04      4.9670E-02    5.4106E-02 7    2.6790E-05    4.1791E-03      1.5495E-05    2.2168E-04      4.9659E-02    5.4102E-02 8    2.6830E-05    4.1854E-03      1.5518E-05    2.2201E-04      4.9649E-02    5.4098E-02 9    2.6867E-05    4.1911E-03      1.5539E-05    2.2232E-04      4.9639E-02    5.4095E-02 10    2.6901E-05    4.1965E-03      1.5559E-05    2.2260E-04      4.9630E-02    5.4092E-02 11    2.6933E-05    4.2015E-03      1.5577E-05    2.2287E-04      4.9622E-02    5.4089E-02 12    2.6963E-05    4.2061E-03      1.5595E-05    2.2311E-04      4.9614E-02    5.4086E-02 13    2.6990E-05    4.2105E-03      1.5611E-05    2.2334E-04      4.9607E-02    5.4083E-02 14    2.7017E-05    4.2145E-03      1.5626E-05    2.2356E-04      4.9600E-02    5.4081E-02 15    2.7077E-05    4.2239E-03      1.5661E-05    2.2406E-04      4.9585E-02    5.4075E-02 16    2.2037E-05    3.4377E-03      1.2746E-05    1.8235E-04      5.0889E-02    5.4544E-02 17    2.2037E-05    3.4377E-03      1.2745E-05    1.8235E-04      5.0889E-02    5.4544E-02 18    1.7683E-05    2.7586E-03      1.0228E-05    1.4633E-04      5.2016E-02    5.4949E-02 19    1.7487E-05    2.7279E-03      1.0114E-05    1.4470E-04      5.2067E-02    5.4967E-02 Table 6.2 Loose Plate Data Number of        Modeled        Modeled Plates to      Number of    U-235 Mass Plate    600 g U-235        Plates      Per Plate (g) 1          22.12              -              -
2          18.47              -              -
3          13.89            14            42.9 4          13.30              -              -
5          10.32            10            60.0 6            9.84              -              -
7            9.43              -              -
8            9.05              9            66.7 9            8.70              -              -
10            8.37              -              -
11            8.07              -              -
12            7.79              -              -
13            7.53              -              -
14            7.28              -              -
15            7.04              7            85.7 16            8.40              -              -
17            8.16              -              -
18            9.99              -              -
19          10.22              -              -
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                Rev. 16, May 2021 Table 6.2 Loose Plate Number Densities (as-modeled)
U-234        U-235      U-236      U-238    Aluminum          Total Plate (atom/b-cm)  (atom/b-cm) (atom/b-cm) (atom/b-cm) (atom/b-cm)    (atom/b-cm) 3    2.1539E-05  3.3600E-03  1.2458E-05  1.7823E-04  5.1018E-02    5.4591E-02 5    2.7492E-05  4.2887E-03  1.5901E-05  2.2749E-04  4.9477E-02    5.4037E-02 8    2.6975E-05  4.2081E-03  1.5602E-05  2.2322E-04  4.9611E-02    5.4085E-02 15    2.7249E-05  4.2508E-03  1.5760E-05  2.2548E-04  4.9540E-02    5.4059E-02 6-12
 
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                            Rev. 16, May 2021 Figure 6.2 ATR Fuel Element Dimensions 6-14
 
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                            Rev. 16, May 2021 Fuel Plate 19 Fuel Plate 1 Fuel Meat Figure 6.2 Fuel Element Model 6-16
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                          Rev. 16, May 2021 6.3 General Considerations Criticality calculations for the ATR FFSC are performed using the three-dimensional Monte Carlo computer code MCNP52. Descriptions of the fuel assembly geometric models are given in Section 6.3.1, Model Configuration. The material properties for all materials used in the models are provided in Section 6.3.2, Material Properties. The computer code and cross section libraries used are provided in Section 6.3.3, Computer Codes and Cross-Section Libraries. Finally, the most reactive configuration is provided in Section 6.3.4, Demonstration of Maximum Reactivity.
6.3.1 Model Configuration Models are developed for both the fuel element and loose plate basket payloads.
6.3.1.1 Fuel Element Payload The model configuration is relatively simple. Most packaging details are conservatively ignored, particularly at the ends. Because the package is long and narrow, array configurations will stack only in the lateral directions (e.g., 5x5x1). Therefore, the end details, for both the package and the fuel element, are conservatively ignored external to the active fuel region, and these end regions are simply modeled as full-density water.
The package consists of two primary structural components, a circular inner tube and a square outer tube, as shown in Appendix 1.3.2, Packaging General Arrangement Drawings. The inner tube has a nominal outer diameter of 6-in and a nominal thickness of 0.12-in. The outer tube has a nominal outer dimension of 8-in and a nominal thickness of 0.188-in. A layer of insulating material 1-in thick is wrapped around the inner tube.
For the inner tube, tolerances are based upon ASTM A2693. The tolerance on the outer diameter (OD) is +/- 0.030-in, and the tolerance on the wall thickness is +/-10%. Tolerances are selected to minimize the spacing between the fuel elements in the array configuration. This spacing is minimized using the maximum OD and minimum wall thickness. Using the minimum wall thickness also reduces parasitic neutron absorption in the steel. Therefore, the modeled tube OD is 6.03-in, the modeled wall thickness is 0.108-in, and the modeled tube ID is 5.814-in.
For the outer tube, the wall thickness tolerance is +/-10% based upon ASTM A5544 (the tolerance for the optional use of ASTM A2405 also falls within this value). Using the minimum wall thickness of 0.169-in reduces parasitic neutron absorption in the steel. Reactivity in the array cases is maximized by minimizing the outer dimensions of the square. A bounding tolerance of 0.1-in is assumed for this dimension based on drawing tolerance in Appendix 1.3.2, Packaging General Arrangement Drawings, for a modeled OD of 7.9-in. The as-fabricated packages will meet this tolerance.
2 MCNP5, MCNP - A General Monte Carlo N-Particle Transport Code, Version 5; Volume II: Users Guide, LA-CP-03-0245, Los Alamos National Laboratory, April, 2003.
3 ASTM A269-02a, Standard Specification for Seamless and Welded Austenitic Stainless Steel Tubing for General Service.
4 ASTM A554-03, Standard Specification for Welded Stainless Steel Mechanical Tubing.
5 ASTM A240-03, Standard Specification for Chromium and Chromium-Nickel Stainless Steel Plate, Sheet, and Strip for Pressure Vessels and for General Applications.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 In the NCT single package models, the inner tube, insulation, and outer tube are modeled explicitly, as shown in Figure 6.3-1 and Figure 6.3-2. Although negligible water ingress is expected during NCT, the inner cavity of the package is assumed to be flooded with water because the package lid does not contain a seal. However, the region between the insulation and the outer tube will remain dry because water cannot enter this region. The Fuel Handling Enclosure (FHE) is conservatively ignored. Modeling the FHE would decrease water reflection in the single package model. However, the neoprene along the sides of the FHE is modeled explicitly using a thickness of 1/8-in. Because neoprene will reduce the reactivity due to parasitic absorption in chlorine, chlorine is removed from the neoprene, and the density is reduced accordingly. In the model, the fuel element is conservatively positioned at the radial center of the inner tube to maximize neutron reflection. The package is reflected with 12-in of full-density water.
The HAC single package model is essentially the same as the NCT single package model.
Damage in the drop tests was shown to be negligible and concentrated at the ends of the package (See Section 2.12.1). As the ends of the package are not modeled, this end damage does not affect the modeling. The various side drops resulted in only minor localized damage to the outer tube, and no observable bulk deformation of the package. Therefore, the minor damage observed will not impact the reactivity. The insulation is replaced with full-density water, and the region between the insulation and outer tube is also filled with full-density water (see Figure 6.3-3). The treatment of the FHE is the same as the NCT single package model. Cases are developed both with and without the FHE neoprene, and with and without chlorine in the neoprene.
As a result of the drop tests, limited damage to the fuel element was observed. The bottom end box sheared off from the main body, although this condition has no effect on the criticality models because the fuel element is not modeled beyond the active fuel region. Limited damage to the fuel element plates was observed at the ends, although this damage is over a short length in a region of low reactivity worth. Slight localized buckling of the fuel plates was also observed in the region of the fuel element side plate vent openings, as the fuel plates are not as well supported in these regions. Because the observed fuel element damage is minor and will have only a negligible effect on reactivity, no damaged fuel element models are developed.
In the NCT array models, a 9x9x1 array is utilized. Although the FHE would survive NCT events with no damage, the FHE is conservatively ignored and the fuel elements are pushed toward the center of the array. Because the fuel elements are transported in a thin (~0.01-in) plastic bag, this plastic bag is assumed to act as a boundary for partial moderation effects. The plastic bag is not modeled explicitly, because it is too thin to have an appreciable effect on the reactivity. Therefore, it is postulated that the fuel element channels may fill with full-density water, while the region between the fuel element and inner tube fills with variable density water.
The partial moderation effects that could be achieved by modeling the FHE explicitly are essentially addressed by the partial moderation analysis using the plastic bag. Also, modeling the FHE explicitly would result in the fuel elements being significantly pushed apart, which is a less reactive condition. Axial movement of the fuel elements is not considered because axial movement would increase the effective active height of the system and reduce the reactivity due to increased leakage. The presence of chlorine-free neoprene is also considered.
In the HAC array models, a 5x5x1 array is utilized. The HAC array models are essentially the same as the NCT array models, except additional cases are developed to determine the reactivity 6-18
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 effect of allowing variable density water in the region between the inner and outer tubes. Cases are also developed with and without the insulation, and with and without chlorine-free neoprene.
The FHE is conservatively ignored for the reasons stated in the previous paragraph. Because the NCT and HAC models are very similar and the NCT models utilize a larger array, the NCT array models are more reactive than the HAC array models.
The detailed moderation assumptions for the array cases are discussed more fully in Section 6.5, Evaluation of Package Arrays under Normal Conditions of Transport, and Section 6.6, Package Arrays under Hypothetical Accident Conditions.
6.3.1.2 Loose Plate Basket Payload The NCT and HAC single package models are shown in Figure 6.3-4 and Figure 6.3-5, respectively. The NCT and HAC packaging models, including tolerances, are consistent with the values used in the fuel element analysis. The difference is that the aluminum loose plate basket and payload of fuel plates is inserted into the cavity. The loose plate basket does not contain neoprene.
The dimensions of the loose plate basket are provided on the packaging general arrangement drawings. The wall thickness of the basket in the central rectangular region is 0.19 +/- 0.06-in.
The cavity width is 4.5 +/- 0.06-in, and the cavity height is 1.62 +/- 0.06-in. The basket wall thickness is modeled at the minimum thickness of 0.13-in to minimize absorption in the aluminum. The inner dimensions of the basket are modeled at the maximum values of 4.56-in x 1.68-in to maximize the volume available for moderation. The radial supports are neglected in the MCNP models.
In the actual loaded configuration, the loose plates are bundled so that the plates are in close contact, and aluminum dunnage plates are used to fill the void space to prevent lateral movement. In the criticality models, the dunnage plates are conservatively ignored. Modeling the dunnage plates would severely restrict the volume available for water moderation. Because no dunnage plates are modeled, the fuel plates are allowed to arrange in the most reactive geometry, including non-regular pitches. Flat plates are modeled rather than curved plates because flat plates are much simpler to model. It is demonstrated that flat plates and curved plates are neutronically equivalent.
Axial movement of the fuel plates is not considered, because this motion would be negligible and is precluded by the basket design, which has a cavity length of 50.5-in. The fuel plates are approximately 49.5-in long, although only the 48-in active length is modeled.
In the NCT and HAC single package models, the fuel basket is centered in the cavity to maximize water reflection, and all water is at full density to maximize moderation and reflection.
In the NCT array analysis, four different plate types are examined: 3, 5, 8, and 15. Plate type 5 is shown to be the most reactive. A number of both regular and non-regular pitches are utilized in order to find the most reactive condition. Plate type 5 is used in all single package and array analyses.
In the NCT array models, a 9x9x1 array is utilized. Water is assumed to be present inside the cavity at a density that maximizes reactivity. To bound any potential damage to the loose plate basket, the rectangular region of each basket is pushed toward the radial center of the array until 6-19
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                    Rev. 16, May 2021 contact is made with the circular tube. This geometry is not considered credible because the ribs will maintain concentricity between the basket and cavity.
In the HAC array models, a 5x5x1 array is utilized. Water is assumed to be present inside the cavity and between the inner and outer tubes at a density that maximizes reactivity. The detailed moderation assumptions for the array cases are discussed more fully in Section 6.5, Evaluation of Package Arrays under Normal Conditions of Transport, and Section 6.6, Package Arrays under Hypothetical Accident Conditions.
The fuel plates are modeled as undamaged in both the NCT and HAC models. As a result of the drop tests, limited buckling of the fuel plates was observed at the end, although this damage is over a short length in a region of low reactivity worth. Because the observed fuel plate damage is minor and will have only a negligible effect on reactivity, no damaged fuel plate models are developed. Also, any anticipated damage is bounded because the most reactive pitch is modeled for both uniform and non-uniform conditions, and the damaged condition is essentially a subset of the conditions already modeled.
6.3.2 Material Properties The fuel meat compositions are provided in Table 6.2-3 and Table 6.2-5 for the fuel element and loose plates, respectively. The fuel plate cladding is aluminum alloy 6061-0, while the side plates may be either aluminum alloy 6061-T6 or 6061-T651. From a criticality perspective, these alloys are essentially aluminum, and in the MCNP models all aluminum alloy structural materials are modeled as pure aluminum with a density of 2.7 g/cm3. The material properties of the remaining packaging and moderating materials are described in the following paragraphs.
The inner and outer tubes of the package are constructed from stainless steel 304. Although MCNP is used in the calculations, the standard compositions for stainless steel 304 are obtained from the SCALE material library6, which is a standard set accepted for use in criticality analyses.
The stainless steel composition and density utilized in the MCNP models are provided in Table 6.3-1.
The insulation material utilized in the NCT models has a density of 6 pounds per cubic foot (0.096 g/cm3). The insulation is composed of Al2O3 and SiO2 in approximately equal quantities, with small (<1 wt%) quantities of other minor constituents. It is assumed in this analysis that the material is simply 50% Al2O3 and 50% SiO2 by weight and the impurities are neglected.
Insulation material properties are provided in Table 6.3-2.
Neoprene (C4H5Cl) has a density of 1.23 g/cm3, and the chemical composition is provided in Table 6.3-3. Because chlorine is a neutron absorber, for models in which the chlorine has been deleted, a density of 0.737 g/cm3 is utilized.
Water is modeled with a density ranging up to 1.0 g/cm3 and the chemical formula H2O. The S(,) card LWTR.60T is used to simulate hydrogen bound to oxygen in water.
6 Standard Composition Library, NUREG/CR-0200, Rev. 6, Volume 3, Section M8, ORNL/NUREG/CSD-2/V3/R6, September 1998.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                    Rev. 16, May 2021 6.3.3 Computer Codes and Cross-Section Libraries MCNP5 v1.30 is used for the criticality analysis7. All cross sections utilized are at room temperature (293.6 K). The uranium isotopes utilize preliminary ENDF/B-VII cross section data that are considered by Los Alamos National Laboratory to be more accurate than ENDF/B-VI cross sections. ENDF/B-V cross sections are utilized for chromium, nickel, and iron because natural composition ENDF/B-VI cross sections are not available for these elements. The remaining isotopes utilize ENDF/B-VI cross sections. Titles of the cross sections utilized in the models have been extracted from the MCNP output and provided in Table 6.3-4. As discussed in Section 6.3.2, the S(,) card LWTR.60T is also used to simulate hydrogen bound to water.
Cases are run with a minimum 2500 neutrons per generation for 250 generations, skipping the first 50. The 1-sigma uncertainty is approximately 0.001 for most cases.
6.3.4 Demonstration of Maximum Reactivity Fuel Element Payload The reactivities of the NCT and HAC single package cases are small, with ks < 0.5.
The NCT and HAC array cases are similar. For the NCT array, a 9x9x1 array is utilized, while in the HAC array, a smaller 5x5x1 array is utilized. Because negligible damage was observed in the drop tests, the package dimensions are the same between the NCT and HAC models.
Dimensions of both the fuel element and packaging are selected to maximize reactivity, and close-water reflection is utilized. In the fuel element, the fuel meat width and channel width are maximized, as this condition is the most reactive. In both NCT and HAC array cases, flooding with partial moderation is allowed in the central cavity, and the fuel elements are pushed toward the center of the array. In the fuel element models, the FHE is not modeled explicitly because the FHE would increase the fuel element spacing and decrease the reactivity. Any partial moderation effects of the FHE are essentially addressed by the partial moderation analysis for the fuel element itself.
In the NCT array models, insulation is modeled between the inner and outer tubes, while in the HAC array models, this region may have water, void, or insulation. In both sets of models, chlorine-free neoprene is modeled adjacent to the fuel element side plates, although the effect on the reactivity is small. No models in which the neoprene is allowed to decompose and homogeneously mix with the water are developed, as this scenario is already bounded by the variable water density search.
The NCT array is more reactive than the HAC array, primarily because the NCT array is significantly larger. The most reactive case (Case E23) results in a ks = 0.8362, which is below the USL of 0.9209.
Loose Plate Basket Payload The reactivities of the NCT and HAC single package cases are small, with ks < 0.5.
To facilitate model preparation, only four different plate types are examined: 3, 5, 8, and 15. The fuel meat width is maximized in all loose plate models, as this condition has been shown to 7
MCNP5, MCNP - A General Monte Carlo N-Particle Transport Code, Version 5; Volume II: Users Guide, LA-CP-03-0245, Los Alamos National Laboratory, April 2003.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 maximize reactivity. For simplicity, plate types are not mixed in the same model. An optimum pitch search is performed to determine the most reactive condition. Both regular and non-regular pitches are examined. Plate 5 is the most reactive because its small width allows this plate to double stack along the width of the basket, resulting in a higher level of moderation compared to the larger plates. Plates 1 through 4 are smaller than Plate 5, but the low uranium loading of these plates results in a higher number of plates to achieve 600 g U-235, and the larger number of plates results in less moderation. In actual practice, plates of any type may be combined in a single loose plate basket, although random combinations of plates would be less reactive than modeling all plates as type 5.
The actual loose plate basket may accept either flat or curved plates. However, plates are modeled as flat rather than curved to facilitate model preparation. It is demonstrated that flat plates are neutronically equivalent to curved plates.
The array geometry and modeling assumption for the loose plate basket payload are similar to those described above for the fuel element payload. The NCT array is more reactive than the HAC array, primarily because the NCT array is significantly larger. The most reactive NCT configuration is with full-water density between the fuel plates, a water density of 0.5 g/cm3 between the basket and the inner pipe, and void between the insulation and the outer tube. The axial regions beyond the active fuel are modeled as water to maximize reflection. The most reactive case (Case LG5) results in a ks = 0.7747, which is below the USL of 0.9209. Note that the most reactive loose plate basket case is less reactive than the most reactive fuel element payload case.
Table 6.3 SS304 Composition Component                Wt.%
C                      0.08 Si                    1.0 P                    0.045 Cr                    19.0 Mn                      2.0 Fe                  68.375 Ni                    9.5 Density = 7.94 g/cm3 6-22
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                  Rev. 16, May 2021 Table 6.3 Insulation Composition Component                Wt.%
Al                  26.5 Si                  23.4 O                  50.2 Density = 0.096 g/cm3 Table 6.3 Neoprene Composition Component                Wt.%
H                    5.7 C                  54.3 Cl                  40.0 Density = 1.23 g/cm3 Table 6.3 Cross Section Libraries Utilized Isotope/Element                Cross Section Label (from MCNP output) 1001.62c    1-h-1 at 293.6K from endf-vi.8 njoy99.50 6000.66c    6-c-0 at 293.6K from endf-vi.6 njoy99.50 8016.62c    8-o-16 at 293.6K from endf-vi.8 njoy99.50 13027.62c    13-al-27 at 293.6K from endf-vi.8 njoy99.50 14000.60c    14-si-nat from endf/b-vi 15031.66c    15-p-31 at 293.6K from endf-vi.6 njoy99.50 17000.66c    17-cl-0 at 293.6K from endf-vi.0 njoy99.50 24000.50c    njoy 25055.62c    25-mn-55 at 293.6K from endf/b-vi.8 njoy99.50 26000.55c    njoy 28000.50c    njoy 92234.69c    92-u-234 at 293.6K from t16 u234la4 njoy99.50 92235.69c    92-u-235 at 293.6K from t16 u235la9d njoy99.50 92236.69c    92-u-236 at 293.6K from t16 u236la2d njoy99.50 92238.69c    92-u-238 at 293.6K from t16 u238la8h njoy99.50 6-23
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                              Rev. 16, May 2021 12-in water reflector 1-in insulation 6.03-in 1/8-in neoprene 7.9-in Figure 6.3 NCT Single Package Model, Fuel Element (planar view) 6-24
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                            Rev. 16, May 2021 48-in Note that the ends of both the fuel element and package are conservatively treated simply as a water reflector.
Figure 6.3 NCT Single Package Model, Fuel Element (axial view) 6-25
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                            Rev. 16, May 2021 Insulation and void replaced with water.
Figure 6.3 HAC Single Package Model, Fuel Element (planar view) 6-26
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                          Rev. 16, May 2021 0.13-in wall thickness 1.68-in 4.56-in Figure 6.3 NCT Single Package Model, Basket (planar view) 6-27
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                          Rev. 16, May 2021 Case LB3 Figure 6.3 HAC Single Package Model, Basket (planar view) 6-28
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                    Rev. 16, May 2021 6.4 Single Package Evaluation Compliance with the requirements of 10 CFR &sect;71.55 is demonstrated by analyzing an optimally moderated damaged and undamaged, single-unit ATR FFSC. The figures and descriptions provided in Section 6.3.1, Model Configuration, describe the basic geometry of the single-unit models.
6.4.1 Single Package Configuration 6.4.1.1 NCT Configuration 6.4.1.1.1        Fuel Element Payload The geometry of the NCT single package configuration is discussed in Section 6.3.1, Model Configuration. The inner tube is flooded with full-density water. The fuel element geometry is consistent with the most reactive fuel element model, including tolerances, as determined in Section 6.4.1.2.1, Fuel Element Payload Parametric Evaluation. Consistent with the most reactive HAC single package model, neoprene from the FHE is modeled at the sides of the fuel element. Chlorine is conservatively removed from the neoprene because chlorine acts as a poison. The package is reflected with 12-in of water.
Two cases are developed. In Case A1, the modeled channel width is 0.085-in, and in Case A2, the modeled channel width is 0.089-in. A channel width of 0.089-in is the maximum local channel width (0.087-in) with an additional margin of 0.002-in. The larger channel width is achieved by reducing the cladding thickness. Case A2 is more reactive, although the reactivity is low, with ks = 0.42239. This result is below the USL of 0.9209. Results are provided in Table 6.4-1.
6.4.1.1.2        Loose Plate Basket Payload The selection of the bounding fuel plate and development of the various plate arrangements are presented in conjunction with the NCT array analysis in Section 6.5.1.2, Loose Plate Basket Payload. It is determined that Plate 5 may be used as a bounding plate type for criticality purposes. Because the aluminum dunnage has not been credited, the plates are allowed to become arranged in the most reactive configuration within the loose plate container. The most reactive fuel plate arrangements determined in the NCT array analysis are used in the NCT single package analysis. The NCT single package models are reflected with 12-in of water.
The 10 Type 5 plates are modeled as 5 plates of double fuel meat width to allow two plates to be present side by side. The top and bottom plates are in contact with the fuel basket inner surfaces, and the center plate is always in the center of the basket. The two off-center plates are shifted in 0.1-cm increments away from the center plate so that the pitch is non-regular. When the pitch is non-regular, the maximum pitch is given as a max value in the results table.
A figure showing the general NCT model geometry is provided in Figure 6.3-4. Results are provided in Table 6.4-2. Six cases are run, with small variations in the plate arrangement. The maximum reactivity occurs for Case LA4, with ks = 0.40199. This result is below the USL of 0.9209. The pitch of this case is non-regular. The top, center, and bottom plates are centered in the lattice locations with a base pitch of 1.036-cm, while the off-center plates are shifted 0.3-cm from the center plate. Note that the most reactive NCT array case peaks with the off-center 6-29
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                      Rev. 16, May 2021 plates shifted 0.2-cm rather than 0.3-cm, although this difference is most likely due to statistical fluctuation.
6.4.1.2 HAC Configuration 6.4.1.2.1        Fuel Element Payload Parametric Evaluation Prior to development of a single package model, a parametric analysis is performed to determine the impacts of various fuel element tolerances on the reactivity. This parametric analysis considers the effects of a number of parameters, such as fuel meat arc length, fuel meat thickness, channel width, and active fuel length.
Because the ATR fuel element is complex, with 19 unique fuel plates and 19 unique fuel material descriptions, performing this parametric study on an actual fuel element would be cumbersome. Rather, the approach utilized is to perform the parametric study on a system of 19 identical flat plates. This geometry mimics the ATR fuel element to determine trends in the data.
Note that the reactivity of the 19 flat plate model is not identical to the reactivity of an actual ATR fuel element due to geometrical and material differences, although the trends are the same.
The most reactive model variations are then incorporated into the ATR fuel element model.
In the parametric models, 1200 g U-235 is equally distributed between 19 identical flat plates.
The base configuration consists of plates with a fuel meat width of 2.65-in (6.7355 cm; the average nominal meat arc length), active fuel height of 48-in, fuel meat thickness of 0.02-in, fuel cladding thickness of 0.015-in (total plate thickness of 0.050-in), and fuel channel thickness of 0.078-in. The geometry of Case B1 is shown in Figure 6.4-1. A total of 12 parametric models are developed, as summarized below.
Case ID                                      Case Description B1          Base case B2          Increase width of fuel meat by 0.1-in B3          Decrease width of fuel meat by 0.1-in B4          Increase thickness of fuel meat by 0.002-in B5          Decrease thickness of fuel meat by 0.002-in B6          Increase thickness of fuel meat by 0.002-in but decrease the cladding thickness to maintain a nominal plate thickness B7          Decrease thickness of fuel meat by 0.002-in but increase the cladding thickness to maintain a nominal plate thickness B8          Increase water channel thickness to 0.085-in B9          Increase water channel thickness to 0.085-in by reducing the cladding thickness B10        Decrease active fuel length to 47.0-in B11        Reduce cladding thickness to the minimum value of 0.008-in B12        Combine cases B2 and B9 In Cases B2 through B12, each case is identical to the base case B1 with the exception of the changes identified in the table above. The pitch, which is the sum of the plate thickness and 6-30
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                    Rev. 16, May 2021 channel thickness, is treated as a dependant variable and is allowed to vary as the independent parameters are changed. For example, in Case B5, decreasing the thickness of the fuel meat decreases the pitch, although the channel thickness remains constant. The detailed model description of the parametric cases is summarized in Table 6.4-3.
The results of the parametric analysis are summarized in Table 6.4-4. Because the uncertainty in the calculation is ~0.001, a difference of at least 0.002 (2 milli-k, abbreviated mk) between the various cases is required in order to distinguish a real effect from statistical fluctuation. The results indicate a reactivity increase of 4.3 mk for Case B2, when the width of the fuel meat is increased, and a decrease of 5.4 mk for Case B3, when the width of the fuel meat is decreased.
Therefore, reactivity increases when the width of the fuel meat is maximized.
The nominal thickness of the fuel meat is 0.02-in. No tolerance on the fuel meat is defined because the fuel plates are fabricated using a rolling process. A thickness tolerance of 0.002-in
(+/-10%) is assumed for computational purposes. In Cases B4 and B5, the fuel meat thickness is adjusted for constant channel thickness and variable pitch, while for Cases B6 and B7 the fuel meat thickness is adjusted for constant plate thickness and nominal pitch. The reactivity fluctuations are within 2 mk in all four cases, and it is concluded that a nominal fuel meat thickness of 0.02-in is acceptable for modeling purposes.
In Case B8, the water channel thickness is increased to 0.085-in (increase in pitch), while in Case B9 the water channel thickness is increased to the maximum by artificially reducing the cladding thickness (nominal pitch). Both cases B8 and B9 show large reactivity gains of 9.6 and 12.9 mk, respectively, indicating that reactivity increases when the water channel thickness increases.
In Case B10, the active fuel length is reduced to a lower bound value of 47.0-in. The reactivity increase is within statistical fluctuation. It may be inferred that increasing the active fuel length would also result in a reactivity effect within statistical fluctuation.
In Case B11, the cladding is reduced to the minimum value of 0.008-in, and the reactivity increases by 5.5 mk. This reactivity gain is likely due to the more compact geometry, as the pitch reduces considerably. This scenario is not directly applicable to an ATR fuel element because the pitch is fixed by the side plates and such a minimum pitch is not possible.
The only cases that show a statistically significant increase are B2, B8, B9, and B11. In Case B12, the increased fuel meat width of Case B2 and increased channel width of Case B9 are combined. This model geometry bounds Case B8, and Case B11 is incorporated in an approximate manner because the cladding thickness has been reduced to accommodate the larger channel. The reactivity of Case B12 represents an increase of 19.5 mk over base Case B1.
6.4.1.2.2      Fuel Element Payload The geometry of the HAC single package configuration is discussed in Section 6.3.1, Model Configuration. Based on the parametric evaluation, three HAC single package ATR fuel element models are developed in order to verify the trends indicated in the parametric analysis:
(1) Case C1, a nominal (base) model, (2) Case C2, a conservative model with the increased channel width consistent with Case B9, and (3) Case C3, an optimized model with both increased channel width and increased meat arc length. In all three models, the FHE neoprene is ignored and a nominal pitch is utilized (i.e., the centerline radial locations of the 19 plates are the same in each model). Note that in Cases C1 and C2, the fuel number densities are computed using nominal fuel meat arc lengths and thus do not correspond to the values in Table 6.2-3. In 6-31
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 the increased channel width models, the channel width is increased by removing cladding. This approach is highly conservative, because it is unlikely (if not impossible) to maximize the channel width between each plate. In an actual fuel element, maximizing the channel width between two plates would likely minimize the channel width between the next two plates, as the overall plate thickness is held to a rather tight tolerance.
The HAC single package results are provided in Table 6.4-5. As expected from the parametric analysis, Case C2 is more reactive than Case C1 (by 13.7 mk), and Case C3 is more reactive than Case C1 (by 17.2 mk). Therefore, it may be concluded that reactivity is maximized in the ATR fuel element by maximizing the fuel meat arc length and maximizing the channel width between the fuel plates. This optimized fuel element is used in all models using the fuel element payload (including NCT single package, NCT array, and HAC array models).
In Cases C1, C2, and C3, the neoprene of the FHE is ignored and treated as full-density water.
In Cases C4 and C5, the effect of neoprene is evaluated. Neoprene is a hydrocarbon with the chemical formula C4H5Cl. Neoprene is present on the FHE and is used to cushion the fuel element. In Case C4, 1/8-in of neoprene is modeled along the sides of the fuel element (see Figure 6.3-3). The small strips of neoprene above and below the fuel element are neglected because these strips are of insufficient mass to affect the reactivity in any appreciable manner.
Inclusion of the neoprene has a pronounced negative effect on the reactivity, presumably due to absorption in the chlorine. In Case C5, the chlorine is deleted from the neoprene, and the density is reduced accordingly. Eliminating the chlorine from the neoprene may be postulated to be a result of decomposition during a fire, although such a scenario is not credible. Case C5 is slightly more reactive than Case C3, although the effect may simply be statistical fluctuation. It may be concluded that chlorine-free neoprene has a negligible effect on the reactivity.
Because the fuel may be transported inside of a plastic bag, it is conservatively assumed that the water density inside of the inner tube can vary independently of the water density inside of the fuel element. To maximize neutron reflection, full-density water is always modeled inside of the tube external to the fuel element, and the fuel element is centered laterally within the tube. In Cases C6 through C10, Case C5 is run with a range of water densities between the fuel plates, and maximum water density in all other regions of the model. Reactivity drops as the water density is reduced between the fuel plates, indicating that the system is under moderated.
Case C5 is the most reactive case when comparing Cases C1 through C10. In Case C11, Case C5 is rerun with the channel width increased from 0.085-in to 0.089-in. A channel width of 0.089-in is the maximum local channel width (0.087-in) with an additional margin of 0.002-in.
The larger channel width is achieved by reducing the cladding thickness. Case C11 is the most reactive, with ks = 0.45237. This result is below the USL of 0.9209.
6.4.1.2.3      Loose Plate Basket Payload The selection of the bounding fuel plate and development of the various plate arrangements are presented in conjunction with the NCT array analysis in Section 6.5.1.2, Loose Plate Basket Payload. The most reactive fuel plate arrangements determined in the NCT array analysis are used in the HAC single package analysis. This arrangement will also be the most reactive in the HAC single package models because both the NCT and HAC models are flooded and behave in a similar manner.
6-32
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 A figure showing the general HAC model geometry is provided in Figure 6.3-5. Results are provided in Table 6.4-6. Six cases are run, with small variations in the plate arrangement. The maximum reactivity occurs for Case LB3, with ks = 0.43629. This result is below the USL of 0.9209. The pitch of this case is non-regular. The top, center, and bottom plates are centered in the lattice locations with a base pitch of 1.036-cm, while the off-center plates are shifted 0.2-cm from the center plate.
6.4.2 Single Package Results Following are the tabulated results for the single package cases. The most reactive configurations are listed in boldface.
Table 6.4 NCT Single Package Results, Fuel Element NCT Case Moderator Density                                      ks Case ID          Filename            (g/cm3)          keff                    (k+2)
A1            NS_M100              1.0        0.41068      0.00097      0.41262 A2          NS_M100_C89              1.0        0.42021      0.00109      0.42239 Table 6.4 NCT Single Package Results, Loose Plate Basket ks Case ID          Filename          Pitch (cm)        keff                    (k+2)
LA1            NS_N5P52            1.036        0.39898      0.00091      0.40080 LA2            NS_N5P52A          1.136 (max)    0.39847      0.00096      0.40039 LA3            NS_N5P52B          1.236 (max)    0.39856      0.00097      0.40050 LA4            NS_N5P52C          1.336 (max)    0.40007      0.00096      0.40199 LA5            NS_N5P52D          1.436 (max)    0.39881      0.00095      0.40071 LA6            NS_N5P52E          1.491 (max)    0.39751      0.00095      0.39941 6-33
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                          Rev. 16, May 2021 Table 6.4 Parametric Analysis Input Data, Fuel Element Parameter          B1          B2        B3        B4        B5            B6 Fuel Arc (cm)      6.7355      6.9895    6.4815    6.7355    6.7355        6.7355 Meat 0.02        0.02      0.02      0.022      0.018        0.022 thickness (in)
Active fuel 48          48        48        48        48            48 height (in)
Channel (in)      0.078      0.078      0.078      0.078      0.078        0.078 Cladding (in)      0.015      0.015      0.015      0.015      0.015        0.014 Total plate 0.050      0.050      0.050      0.052      0.048        0.050 (in)
Pitch (in)      0.128      0.128      0.128      0.130      0.126        0.128 Volume (cm3)      41.7164    43.2895    40.1432    45.8880    37.5447      45.8880 U-235 (g)        63.2        63.2      63.2      63.2      63.2          63.2 U-235 density 1.51        1.46      1.57      1.38      1.68          1.38 (g/cm3)
UAlx+Al density        3.86        3.81      3.91      3.74      4.00          3.74 (g/cm3)
N U-234      2.4865E-05  2.3962E-05 2.5840E-05 2.2605E-05 2.7628E-05    2.2605E-05 N U-235      3.8789E-03  3.7380E-03 4.0309E-03 3.5263E-03 4.3099E-03    3.5263E-03 N U-236      1.4382E-05  1.3859E-05 1.4945E-05 1.3074E-05 1.5980E-05    1.3074E-05 N U-238      2.0576E-04  1.9828E-04 2.1382E-04 1.8705E-04 2.2862E-04    1.8705E-04 N U-Al      5.0157E-02  5.0391E-02 4.9905E-02 5.0742E-02 4.9442E-02    5.0742E-02 Total      5.4281E-02  5.4365E-02 5.4190E-02 5.4491E-02 5.4024E-02    5.4491E-02 Parameter          B7        B8        B9        B10        B11          B12 Fuel Arc (cm)      6.7355    6.7355    6.7355    6.7355      6.7355        6.9895 Meat 0.018      0.02      0.02      0.02      0.02          0.02 thickness (in)
Active fuel 48        48        48        47        48            48 height (in)
Channel (in)        0.078      0.085      0.085      0.078      0.078        0.085 Cladding (in)      0.016      0.015    0.0115      0.015      0.008        0.0115 Total plate (in)      0.050      0.050    0.0430      0.050      0.036        0.0430 Pitch (in)        0.128      0.135      0.128      0.128      0.114        0.128 Volume (cm3)      37.5447    41.7164    41.7164    40.8473    41.7164      43.2895 U-235 (g)          63.2      63.2      63.2        63.2      63.2          63.2 U-235 density 1.68      1.51      1.51      1.55      1.51          1.46 (g/cm3)
UAlx+Al 4.00      3.86      3.86      3.89      3.86          3.81 density (g/cm3)
N U-234      2.7628E-05 2.4865E-05 2.4865E-05 2.5394E-05 2.4865E-05    2.3962E-05 N U-235      4.3099E-03 3.8789E-03 3.8789E-03 3.9615E-03 3.8789E-03    3.7380E-03 N U-236      1.5980E-05 1.4382E-05 1.4382E-05 1.4688E-05 1.4382E-05    1.3859E-05 N U-238      2.2862E-04 2.0576E-04 2.0576E-04 2.1014E-04 2.0576E-04    1.9828E-04 N U-Al      4.9442E-02 5.0157E-02 5.0157E-02 5.0020E-02  5.0157E-02    5.0391E-02 Total      5.4024E-02 5.4281E-02 5.4281E-02 5.4232E-02  5.4281E-02    5.4365E-02 6-34
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                            Rev. 16, May 2021 Table 6.4 Parametric Analysis Results, Fuel Element ks        from Case ID        Filename        keff            (k+2)    B1 (mk)
B1              P1        0.46601  0.00096  0.46793        --
B2              P2        0.47015  0.00102  0.47219        4.3 B3              P3        0.46045  0.00102  0.46249      -5.4 B4              P5        0.46403  0.00101  0.46605      -1.9 B5              P4        0.46442  0.00111  0.46664      -1.3 B6            P10        0.46753  0.00105  0.46963        1.7 B7              P9        0.46683  0.00101  0.46885      0.9 B8              P6        0.47528  0.00112  0.47752        9.6 B9              P7        0.47879  0.00100  0.48079      12.9 B10              P8        0.46704  0.00106  0.46916        1.2 B11            P11        0.47123  0.00108  0.47339        5.5 B12            P12        0.48534  0.00104  0.48742      19.5 6-35
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                Rev. 16, May 2021 Table 6.4 HAC Single Package Results, Fuel Element Water Density Between Plates                              ks Case ID        Filename          (g/cm3)      keff              (k+2)
C1    HS_M100_NOM                1.0      0.42274  0.00095      0.42464 C2    HS_M100_TOL                1.0      0.43639  0.00099      0.43837 C3    HS_M100_TOLW              1.0      0.43991  0.00097      0.44185 C4    HS_M100_TOLW_N1            1.0      0.41002  0.00102      0.41206 C5    HS_M100_TOLW_N2            1.0      0.44040  0.00104      0.44248 C6    HS_M050                    0.5      0.35396  0.00088      0.35572 C7    HS_M060                    0.6      0.36994  0.00095      0.37184 C8    HS_M070                    0.7      0.38607  0.00099      0.38805 C9    HS_M080                    0.8      0.40411  0.00102      0.40615 C10    HS_M090                    0.9      0.42092  0.00096      0.42284 C11    HS_M100_TOLW_N2_C89        1.0      0.45029  0.00104      0.45237 Table 6.4 HAC Single Package Results, Loose Plate Basket ks Case ID        Filename    Pitch (cm)      keff              (k+2)
LB1        HS_N5P52        1.036    0.43263  0.00097      0.43457 LB2        HS_N5P52A    1.136 (max)  0.43350  0.00092      0.43534 LB3        HS_N5P52B    1.236 (max)  0.43443  0.00093      0.43629 LB4        HS_N5P52C    1.336 (max)  0.43388  0.00096      0.43580 LB5        HS_N5P52D    1.436 (max)  0.43328  0.00091      0.43510 LB6        HS_N5P52E    1.491 (max)  0.43169  0.00089      0.43347 6-36
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                          Rev. 16, May 2021 Figure 6.4 Base Parametric Model (Case B1) 6-37
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                    Rev. 16, May 2021 6.5 Evaluation of Package Arrays under Normal Conditions of Transport 6.5.1 NCT Array Configuration 6.5.1.1 Fuel Element Payload The NCT array model is a 9x9x1 array of the NCT single package model, see Figure 6.5-1.
Although an 8x8x1 array is of sufficient size to justify a CSI = 4.0, the larger 9x9x1 array is utilized simply for modeling convenience. Neoprene is modeled without chlorine in all models.
It is demonstrated in Section 6.6.1.1, Fuel Element Payload, that chlorine-free neoprene may have a slight positive effect on the reactivity, although the effect is small. The entire array is reflected with 12-in of full-density water.
The fuel elements are pushed to the center of the array and rotated to minimize the distance between the fuel elements. This geometry is not feasible for NCT, because the FHE would force the fuel elements to remain in the center of the package, although the FHE does allow rotation.
Therefore, it is conservative to ignore the FHE to minimize the separation distance. In addition, a small notch is added to the neoprene so that the fuel element may be translated to the maximum extent without interfering with the inner tube geometry. This notch is not present in the single package models.
Three calculational series are developed. In Series 1, the water density is fixed at 1.0 g/cm3 between the fuel plates and the water density is allowed to vary inside the inner tube. The channel width is modeled at 0.085-in. Series 2 is the same as Series 1, although the density within the fuel plates is at a reduced density of 0.9 g/cm3. Void is always present between the insulation and the outer tube, as this region is water-tight. Series 3 is a repeat of Series 1, although with the channel width increased to 0.089-in. A channel width of 0.089-in is the maximum local channel width (0.087-in) with an additional margin of 0.002-in. The larger channel width is achieved by reducing the cladding thickness. The results are provided in Table 6.5-1.
Reactivity is at a maximum for Case E23, which has full-density water between the fuel plates, and 0.3 g/cm3 water inside the inner tube, and a channel width of 0.089-in, with ks = 0.83616.
As expected, the reactivity drops when the water density between the fuel plates is reduced, as the system is under moderated. The maximum result is far below the USL of 0.9209.
As a point of interest, an additional case (Case D12) is developed in which the fuel elements are centered in the cavity and not rotated, using the moderation and channel width assumptions of Case D4 (see the lower figure of Figure 6.5-1). The reactivity drops by 18.5 mk, which essentially represents the additional conservatism of pushing the fuel elements to the center of the array.
6.5.1.2 Loose Plate Basket Payload The NCT array model is a 9x9x1 array of the NCT single package model. For the NCT single package cases, it was sufficient to laterally center the fuel basket within the inner tube to maximize reflection by the water in the tube. However, in the NCT array configuration, it is expected that reactivity would be maximized by pushing the fuel baskets to the center of the 6-38
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                    Rev. 16, May 2021 array, as shown in the top sketch of Figure 6.5-2. The fuel elements may be packed closer by rotating them as shown in the figure. Therefore, unless otherwise noted, all NCT array models have the baskets pushed toward the center of the array. Although this assumption bounds any anticipated basket damage, this arrangement is not credible, because the structural ribs that center the baskets within the inner tube will not deform in this manner.
The loose plate payload consists of 19 different plate types. Each plate type has a different width and uranium mass, although the lengths are the same. Each plate may be either flat or curved, for a total of 19*2 = 38 different variations. However, flat and curved plates will not be mixed in the same basket (to facilitate packaging). Within each loose plate basket, any combination of plate types may be present, with the only limitation that the total U-235 mass present in the basket must not exceed 600 g.
Clearly, there are a large number of possible combinations of plates that may be present within the basket. The objective is to determine a simplified configuration that bounds any random collection of plates. Fortunately, calculations may be performed using only flat plates, because the neutronic behavior of flat and curved plates is demonstrated to be nearly identical.
Therefore, the flat plate results also apply to the curved plates. Flat plates allow easy geometry setup using MCNP repeated structures, while curved plates generally cannot be modeled using repeated structures unless the plate pitch is rather large.
Basic data for the 19 plate types are provided in Table 6.2-1. It is not necessary to model each of the 19 different plate types. Rather, from examination of these data, a subset of plates is selected for further analysis. Plates 5 through 15 have a U-235 density of approximately 1.64 g/cm3, while the remaining plates have a significantly lower U-235 density. Plate 5 is the smallest plate in this range, and Plate 15 is the largest plate in this range; both are selected for further evaluation. Plate 8 is also selected as a representative plate between these two extremes, and should result in reactivity values between Plate 5 and 15. It is demonstrated that the smaller plate configuration (Plate 5) is more reactive than the larger plate configurations (Plates 8 and 15). Plate 3 is also selected for further evaluation because it is smaller than Plate 5, although the reduced U-235 density will result in a larger number of plates.
For simplicity, only one plate type is modeled within each basket. Randomly mixing different plate types would result in a less reactive condition that the most reactive single plate configuration. Also, number densities of the selected plates have been slightly adjusted so that the total mass of U-235 is always 600 g. For plates 5, 8, and 15, the number densities are increased, while for plate 3 the number densities are decreased.
Four initial series of calculations are performed, one series for each of the four plate types under consideration. The goal of these initial calculations is to simply determine the bounding plate type. Once the bounding plate type has been determined, additional series of calculations are performed on the bounding plate type. For all of the initial models, full-density water is modeled between the plates, 0.3 g/cm3 water is modeled between the plate array and basket (this region is not present once the plate array fills the entire basket area), 0.3 g/cm3 water is modeled between the basket and inner tube, and insulation/void is modeled between the inner and outer tubes. The water density of 0.3 g/cm3 is selected based upon the most reactive moderation condition of the ATR fuel element analysis, and will be optimized once the bounding plate is selected.
Fuel Plate 5 Series: Fuel plate 5 is the first plate type examined. Ten plates are required to achieve a mass of 600 g U-235. The plate arrangements for a number of the configurations are 6-39
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                    Rev. 16, May 2021 shown in Figure 6.5-3 through Figure 6.5-5. Results are provided in Table 6.5-2. In cases LC1 through LC9, the plates are arranged in a 1x10 array at the center of the basket. Reactivity is low when the pitch is small, and reactivity increases as the pitch increases. In cases LC10 and LC11, the reactivity increases as the plates are alternately shifted to the right and left because moderation increases. In case LC12, plates are alternately shifted up and down until they contact each other.
Because fuel plate 5 is rather narrow, it is possible to further increase the moderation by modeling the plates in a 2x5 array in cases LC13 through LC29. Because the plate is slightly too wide to fit two side-by-side, the two side-by-side plates are modeled as a single plate by doubling the fuel meat width. The reactivity continues to increase with increasing pitch. Case LC19 has the largest reactivity obtained with a constant pitch.
However, moderation can be further increased if a non-regular pitch is utilized. In cases LC20 through LC29, non-regular pitches are examined. In these cases, the plates at the top, center, and bottom of the basket remain fixed, while the two off-center plates are shifted away from the center plate in 0.1-cm increments. Because the pitches in these cases are non-regular, the pitches provided in the results table are noted as max values. Case LC21 is the most reactive, with ks=0.76806, although the reactivity gain resulting from a non-regular pitch is relatively small and within statistical fluctuation. For case LC21, the top, center, and bottom plates are centered in the lattice locations with a base pitch of 1.036 cm, while the off-center plates are shifted 0.2-cm from the center plate (maximum pitch of 1.236 cm).
Fuel Plate 8 Series: Nine plates are required to achieve a mass of 600 g U-235. The plate arrangements for a number of configurations are shown in Figure 6.5-6 and Figure 6.5-7.
Results are provided in Table 6.5-3. Considerably fewer cases are generated compared to fuel plate 5 because it has been established that the plates are highly under moderated when packed tightly.
In cases LD1 through LD3, the plates are modeled in a simple 1x9 array. In cases LD4 though LD11, the plates are alternately shifted left and right to increase moderation. In cases LD6 through LD11, the top, bottom, and center plates remain fixed, while the remaining plates are progressively shifted up or down in 0.1-cm increments. Case LD7 is the most reactive, with ks=0.75241, although the reactivity is less than the most reactive plate 5 case. For case LC7, the base lattice pitch is 0.574-cm, and the off-center plates are shifted 0.2-cm from the center plate.
Fuel Plate 15 Series: Seven plates are required to achieve a mass of 600 g U-235. The plate arrangements for a number of the configurations are shown in Figure 6.5-8. Results are provided in Table 6.5-3. Using the same methodology as plates 5 and 8, case LE8 is the most reactive, with ks=0.74548. This case also features a non-regular pitch. For case LE8, the base lattice pitch is 0.804 cm, and the off-center plates are shifted 0.1 cm from the center plate.
Comparing the maximum ks values for plates 5, 8, and 15, plate 5 is the most reactive (ks=0.76806), plate 15 is the least reactive (ks=0.74548), and plate 8 falls between the two (ks=0.75241). In fact, the reactivities of plates 8 and 15 are fairly close, despite the difference in the width and number of plates. Plate 5 is somewhat more reactive than either plate 8 or 15, most likely because its narrow width allows double stacking of this plate along the width of the basket, which results in a more advantageous moderation and geometry conditions. Therefore, the trend is that for a fixed U-235 mass per basket, the smaller plates are more reactive than the larger plates.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                    Rev. 16, May 2021 Of course, plates 1 through 4 are smaller than plate 5. However, these plates have a lower U-235 density so that more plates are required to achieve 600 g U-235. More plates would provide less volume for moderation, so it is expected that plate 5 would bound plates 1 through 4. This is confirmed by running several cases for plate 3.
Fuel Plate 3 Series: Fourteen plates are required to achieve a mass of 600 g U-235. The plate arrangements for the configurations are shown in Figure 6.5-9. Results are provided in Table 6.5-3. All cases are for a 2x7 arrangement and non-regular pitches, as similar arrangements have been shown to be the most reactive for the other plates. Two side-by-side plates are modeled as a single plate with double fuel meat width, consistent with the treatment of the Type 5 plate.
Case LF2 is the most reactive, with ks=0.75904, although this case is less reactive than the Type 5 plate. For case LF2, the pitch is 0.796 cm.
Criticality Analysis Using Plate 5: From the analysis of plate types 3, 5, 8, and 15, Type 5 is shown to be the most reactive. Therefore, the remaining analysis uses only this plate type. An additional two series of cases are performed using fuel plate 5 in which the water densities in the various model regions are allowed to vary. The primary regions of interest are within the basket and between the basket and the inner tube.
In Series 1, full-density water is modeled within the basket, while the water density between the basket and the inner tube is varied from 0 to 1.0 g/cm3. The results are provided in Table 6.5-4.
The maximum reactivity occurs for Case LG5, with ks = 0.77469. A water density of 0.5 g/cm3 within the inner tube is utilized in the most reactive case.
In Series 2, the water density inside the basket is reduced to 0.9 g/cm3, while the water density between the basket and the inner tube is varied from 0 to 1.0 g/cm3. The reactivity clearly drops when reduced density water is modeled inside the basket.
Several miscellaneous cases are run to validate the assumptions noted above. In Case LJ1, the most reactive case (Case LG5) is run with the fuel baskets centered inside of the tubes (see the lower sketch of Figure 6.5-2). The reactivity drops as the fuel elements are pushed apart, ks =
0.76237 for Case LJ1, compared to ks = 0.77469 for Case LG5.
It has been implicitly assumed the maximum reactivity is obtained for the maximum fissile mass of 600 g U-235. In general, the maximum allowable fissile loading is not necessarily the most reactive condition if the volume of fissile material is so large that little volume is available for moderating material. That is not the case for the loose plate analysis, as the fuel plates are thin and only a small number of plates are required to achieve a mass of 600 g U-235. Removing plates might increase moderation slightly as water is added to the system, although reducing the fissile mass more than compensates for the additional moderation and lowers the reactivity. To demonstrate this effect, the arrangement of Case LC9, which has ten type 5 plates in a 1x10 evenly spaced array (see Figure 6.5-3), is repeated with ten, nine, eight, and seven evenly spaced plates (Cases LJ2, LJ3, LJ4, and LJ5, see Figure 6.5-10) with an inner tube water density of 0.5 g/cm3. The reactivity drops as each successive plate is removed (0.62333 for Case LJ2 to 0.57579 for Case LJ5), despite the fact that the plates are spaced farther and farther apart and moderation is improved. If plates are removed from the most reactive models, for which the pitch is already non-regular to maximize reactivity, the reactivity drop resulting from removing plates would be more pronounced.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 It is stated that modeling the plates as flat is neutronically equivalent to modeling the plates as curved. This modeling assumption is verified by modeling both flat and curved plates with a constant pitch of 0.80 cm. This pitch is selected because it is large and constant and the curved plates may be modeled with repeated structures. Case LJ6 is the flat plate model, and Case LJ7 is the curved plate model. Case LJ6 is geometrically identical to case LC13 (see Figure 6.5-4) except the water density inside the basket is 1.0 g/cm3 between the plate array and the basket.
Case LJ7 is shown in Figure 6.5-10. Flat plate Case LJ6 has ks=0.73021, while curved plate Case LJ7 has ks=0.73022. The difference between these cases is negligible, and the statement that flat plates are neutronically equivalent to curved plates is verified.
In conclusion, Case LG5 is the most reactive loose plate basket model, with ks = 0.77469. This result is below the USL of 0.9209. Case LG5 has fully moderated fuel plates, 0.5 g/cm3 water inside the inner tube, and fuel plate baskets that have been rotated and moved to the center of the array.
6.5.2 NCT Array Results The results for the NCT array cases are provided in the following table. The most reactive configuration is listed in boldface.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 Table 6.5 NCT Array Results, Fuel Element Payload Water        Water        Water Density        Density      Density Between      Inside Inner  Between Case                            Tubes          Tube        Plates                            ks ID          Filename        (g/cm3)      (g/cm3)      (g/cm3)      keff            (k+2)
Series 1: Variable water density inside inner tube, full density water between plates.
D1          NA_P000            0              0            1.0      0.76716  0.00120  0.76956 D2          NA_P010            0            0.1          1.0      0.80349  0.00123  0.80595 D3          NA_P020            0            0.2          1.0      0.81928  0.00112  0.82152 D4          NA_P030            0            0.3          1.0      0.82605  0.00117  0.82839 D5          NA_P040            0            0.4          1.0      0.82149  0.00119  0.82387 D6          NA_P050            0            0.5          1.0      0.81420  0.00118  0.81656 D7          NA_P060            0            0.6          1.0      0.80521  0.00108  0.80737 D8          NA_P070            0            0.7          1.0      0.79216  0.00121  0.79458 D9          NA_P080            0            0.8          1.0      0.78130  0.00132  0.78394 D10          NA_P090            0            0.9          1.0      0.76905  0.00120  0.77145 D11          NA_P100            0            1.0          1.0      0.75603  0.00124  0.75851 D12          NA_P030C            0            0.3          1.0      0.80743  0.00122  0.80987 Series 2: Variable water density inside inner tube, 0.9 g/cm3 density water between plates.
E1        NA_M90P000            0              0            0.9      0.72938  0.00111  0.73160 E2        NA_M90P010            0            0.1          0.9      0.77108  0.00120  0.77348 E3        NA_M90P020            0            0.2          0.9      0.79299  0.00116  0.79531 E4        NA_M90P030            0            0.3          0.9      0.79943  0.00123  0.80189 E5        NA_M90P040            0            0.4          0.9      0.80192  0.00108  0.80408 E6        NA_M90P050            0            0.5          0.9      0.79378  0.00108  0.79594 E7        NA_M90P060            0            0.6          0.9      0.78539  0.00111  0.78761 E8        NA_M90P070            0            0.7          0.9      0.77658  0.00118  0.77894 E9        NA_M90P080            0            0.8          0.9      0.76496  0.00117  0.76730 E10        NA_M90P090            0            0.9          0.9      0.75315  0.00121  0.75557 E11        NA_M90P100            0            1.0          0.9      0.74334  0.00126  0.74586 Series 3: Variable water density inside inner tube, full density water between plates, channel width increased to 0.089-in E20      NA_P000_C89            0              0            1.0      0.78147  0.00115  0.78377 E21      NA_P010_C89            0            0.1          1.0      0.81601  0.00119  0.81839 E22      NA_P020_C89            0            0.2          1.0      0.83027  0.00119  0.83265 E23        NA_P030_C89            0            0.3          1.0      0.83372  0.00122  0.83616 E24      NA_P040_C89            0            0.4          1.0      0.82984  0.00118  0.83220 E25      NA_P050_C89            0            0.5          1.0      0.82481  0.00125  0.82731 E26      NA_P060_C89            0            0.6          1.0      0.81462  0.00119  0.81700 6-43
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                              Rev. 16, May 2021 Table 6.5 NCT Array Results, Pitch Variations, Plate 5 ks Case ID      Filename  Pitch (cm)    keff          (k+2)
LC1        NA_N5P08    0.160    0.41955 0.00093  0.42141 LC2        NA_N5P10    0.200    0.44783 0.00091  0.44965 LC3        NA_N5P12    0.240    0.47653 0.00104  0.47861 LC4        NA_N5P14    0.280    0.50372 0.00104  0.50580 LC5        NA_N5P16    0.320    0.53109 0.00103  0.53315 LC6        NA_N5P18    0.360    0.55470 0.00109  0.55688 LC7        NA_N5P20    0.400    0.57669 0.00104  0.57877 LC8        NA_N5P22    0.440    0.59930 0.00111  0.60152 LC9        NA_N5P23    0.460    0.61120 0.00102  0.61324 LC10      NA_N5P23A    0.460    0.69108 0.00118  0.69344 LC11      NA_N5P23B    0.460    0.74866 0.00109  0.75084 LC12      NA_N5P23C    0.460    0.74714 0.00102  0.74918 LC13        NA_N5P40    0.800    0.71462 0.00107  0.71676 LC14        NA_N5P42    0.840    0.72319 0.00108  0.72535 LC15        NA_N5P44    0.880    0.73353 0.00102  0.73557 LC16        NA_N5P46    0.920    0.74169 0.00107  0.74383 LC17        NA_N5P48    0.960    0.74962 0.00112  0.75186 LC18        NA_N5P50    1.000    0.75920 0.00109  0.76138 LC19        NA_N5P52    1.036    0.76423 0.00118  0.76659 LC20      NA_N5P52A  1.136 (max)  0.76520 0.00102  0.76724 LC21      NA_N5P52B  1.236 (max)  0.76582 0.00112  0.76806 LC22      NA_N5P52C  1.336 (max)  0.76393 0.00107  0.76607 LC23      NA_N5P52D  1.436 (max)  0.76254 0.00096  0.76446 LC24      NA_N5P52E  1.493 (max)  0.75949 0.00093  0.76135 LC25        NA_N5P67  1.540 (max)  0.75942 0.00101  0.76144 LC26      NA_N5P67A  1.640 (max)  0.75508 0.00105  0.75718 LC27      NA_N5P67B  1.740 (max)  0.74803 0.00106  0.75015 LC28      NA_N5P67C  1.840 (max)  0.73839 0.00107  0.74053 LC29      NA_N5P67D  1.940 (max)  0.72412 0.00105  0.72622 6-44
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                              Rev. 16, May 2021 Table 6.5 NCT Array Results, Pitch Variations, Plates 8, 15, and 3 ks Case ID      Filename  Pitch (cm)    keff          (k+2)
Plate 8 LD1        NA_N8P22      0.440    0.60412  0.00106  0.60624 LD2        NA_N8P24      0.480    0.62588  0.00106  0.62800 LD3        NA_N8P26      0.518    0.64309  0.00112  0.64533 LD4      NA_N8P26A      0.518    0.74015  0.00102  0.74219 LD5        NA_N8P29      0.574    0.74719  0.00105  0.74929 LD6      NA_N8P29A  0.674 (max) 0.74875  0.00112  0.75099 LD7      NA_N8P29B  0.774 (max) 0.75035  0.00103  0.75241 LD8        NA_N8P29C  0.874 (max) 0.74896  0.00099  0.75094 LD9      NA_N8P29D  0.974 (max) 0.74574  0.00102  0.74778 LD10      NA_N8P29E  1.074 (max) 0.74373  0.00092  0.74557 LD11      NA_N8P29F  1.174 (max) 0.73494  0.00106  0.73706 Plate 15 LE1      NA_N15P32    0.640    0.68653  0.00107  0.68867 LE2      NA_N15P34    0.690    0.70200  0.00113  0.70426 LE3      NA_N15P34A    0.690    0.73590  0.00114  0.73818 LE4      NA_N15P34B  0.790 (max) 0.74090  0.00110  0.74310 LE5      NA_N15P34C  0.890 (max) 0.74003  0.00111  0.74225 LE6      NA_N15P34D  0.970 (max) 0.74209  0.00108  0.74425 LE7      NA_N15P40    0.804    0.74153  0.00115  0.74383 LE8      NA_N15P40A  0.904 (max) 0.74322  0.00113  0.74548 LE9      NA_N15P40B  1.004 (max) 0.74089  0.00118  0.74325 LE10      NA_N15P40C  1.104 (max) 0.73801  0.00100  0.74001 Plate 3 LF1      NA_N3P40      0.796    0.75062  0.00102  0.75266 LF2      NA_N3P40A      0.796    0.75696  0.00104  0.75904 LF3      NA_N3P40B  0.896 (max) 0.75655  0.00107  0.75869 LF4      NA_N3P40C  0.996 (max) 0.75365  0.00094  0.75553 LF5      NA_N3P40D  1.096 (max) 0.75155  0.00106  0.75367 6-45
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                Rev. 16, May 2021 Table 6.5 NCT Array Results, Plate 5 Water          Water          Water Density        Density        Density Between      Inside Inner    Between Case                          Tubes          Tube          Plates                            ks ID      Filename          (g/cm3)        (g/cm3)        (g/cm3)        keff            (k+2)
Series 1: Variable water density inside inner tube, full-density water in basket.
LG1      NA_N5P000              0              0              1.0      0.66797  0.00097  0.66991 LG2      NA_N5P010              0              0.1            1.0      0.71859  0.00100  0.72059 LG3      NA_N5P020              0              0.2            1.0      0.74925  0.00104  0.75133 LC21    NA_N5P52B              0              0.3            1.0      0.76582  0.00112  0.76806 LG4      NA_N5P040              0              0.4            1.0      0.77225  0.00117  0.77459 LG5      NA_N5P050              0              0.5            1.0      0.77251  0.00109  0.77469 LG6      NA_N5P060              0              0.6            1.0      0.76738  0.00099  0.76936 LG7      NA_N5P070              0              0.7            1.0      0.75998  0.00100  0.76198 LG8      NA_N5P080              0              0.8            1.0      0.75086  0.00114  0.75314 LG9      NA_N5P090              0              0.9            1.0      0.74066  0.00111  0.74288 LG10    NA_N5P100              0              1.0            1.0      0.72764  0.00111  0.72986 Series 2: Variable water density inside inner tube, reduced density water in basket.
LH1  NA_N5M090P000            0              0              0.9      0.63496  0.00098  0.63692 LH2  NA_N5M090P010            0              0.1            0.9      0.69390  0.00093  0.69576 LH3  NA_N5M090P020            0              0.2            0.9      0.72793  0.00095  0.72983 LH4  NA_N5M090P030            0              0.3            0.9      0.74560  0.00108  0.74776 LH5  NA_N5M090P040            0              0.4            0.9      0.75402  0.00108  0.75618 LH6  NA_N5M090P050            0              0.5            0.9      0.75480  0.00109  0.75698 LH7  NA_N5M090P060            0              0.6            0.9      0.75429  0.00110  0.75649 LH8  NA_N5M090P070            0              0.7            0.9      0.74414  0.00100  0.74614 LH9  NA_N5M090P080            0              0.8            0.9      0.73639  0.00104  0.73847 LH10  NA_N5M090P090            0              0.9            0.9      0.72573  0.00095  0.72763 LH11  NA_N5M090P100            0              1.0            0.9      0.71549  0.00107  0.71763 Miscellaneous Cases LJ1    NA_N5P050C              0              0.5            1.0      0.76003  0.00117  0.76237 LJ2    NA_N5P23_10            0              0.5            1.0      0.62119  0.00107  0.62333 LJ3    NA_N5P23_9              0              0.5            1.0      0.60657  0.00106  0.60869 LJ4    NA_N5P23_8              0              0.5            1.0      0.59251  0.00114  0.59479 LJ5    NA_N5P23_7              0              0.5            1.0      0.57369  0.00105  0.57579 LJ6    NA_N5P40_F              0              0.5            1.0      0.72815  0.00103  0.73021 LJ7    NA_N5P40_C              0              0.5            1.0      0.72810  0.00106  0.73022 6-46
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                          Rev. 16, May 2021 Pushed to center of array Centered in each tube (D12)
Figure 6.5 NCT Array Geometry, Fuel Element Payload 6-47
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                          Rev. 16, May 2021 Pushed to center of array Centered in each tube (LJ1)
Figure 6.5 NCT Array Geometry, Loose Plate Basket Payload 6-48
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                        Rev. 16, May 2021 LC1 LC9 LC10 LC11 Figure 6.5 NCT Array Geometry, Plate 5 (LC1, LC9, LC10, LC11) 6-49
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                        Rev. 16, May 2021 LC12 LC13 LC19 LC21 Figure 6.5 NCT Array Geometry, Plate 5 (LC12, LC13, LC19, LC21) 6-50
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                        Rev. 16, May 2021 LC22 LC24 LC26 LC29 Figure 6.5 NCT Array Geometry, Plate 5 (LC22, LC24, LC26, LC29) 6-51
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                          Rev. 16, May 2021 LD3 LD4 LD5 LD7 Figure 6.5 NCT Array Geometry, Plate 8 (LD3, LD4, LD5, LD7) 6-52
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                        Rev. 16, May 2021 LD8 LD9 LD10 LD11 Figure 6.5 NCT Array Geometry, Plate 8 (LD8, LD9, LD10, LD11) 6-53
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                          Rev. 16, May 2021 LE2 LE3 LE6 LE8 Figure 6.5 NCT Array Geometry, Plate 15 (LE2, LE3, LE6, LE8) 6-54
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                          Rev. 16, May 2021 LF1 LF2 LF4 LF5 Figure 6.5 NCT Array Geometry, Plate 3 (LF1, LF2, LF4, LF5) 6-55
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                          Rev. 16, May 2021 LJ3 LJ4 LJ5 LJ7 Figure 6.5 NCT Array Geometry, Miscellaneous (LJ3, LJ4, LJ5, LJ7) 6-56
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 6.6 Package Arrays under Hypothetical Accident Conditions 6.6.1 HAC Array Configuration 6.6.1.1 Fuel Element Payload The HAC array model is a 5x5x1 array of the HAC single package model. As the FHE is assumed to be damaged, the fuel is free to move laterally within the package. To minimize the distance between the fuel elements, in all HAC array models the fuel elements are rotated and moved toward the center of the array, consistent with the NCT array configuration. The FHE is not modeled, because modeling the FHE in any capacity would push the fuel elements apart and lower the reactivity.
From the HAC single package analysis and NCT array analysis, it is known that reactivity is maximized with full-density water between the fuel plates, because the fuel elements are under moderated. Therefore, all HAC array models have full-density water between the fuel plates.
Because the fuel elements may be transported in a plastic bag, it is assumed that the water density between the plates may vary independently from the water density inside the inner tube.
This partial moderation effect is similar to the partial moderation effect that could be achieved by modeling the FHE explicitly.
Nine computational series are performed. The variables addressed are (1) water density inside inner tube, (2) water density between tubes, (3) presence of insulation, and (4) presence of FHE neoprene. The geometries of two of these series are shown in Figure 6.6-1, and the geometries of the other configurations are similar. These nine computational series are described in the following paragraphs. The full results are provided in Table 6.6-1.
In Series 1, the water density inside the inner tube is varied from 0 to 1.0 g/cm3, while void is modeled between the tubes. The modeled channel width is 0.085-in. The insulation and FHE neoprene are not modeled. The maximum reactivity occurs for Case F9, with ks = 0.72933. A water density of 0.8 g/cm3 within the inner tube is utilized in the most reactive case.
In Series 2, the most reactive case from Series 1 (Case F9) is modified so that the water density between the tubes is varied between 0 and 1.0 g/cm3, while the water density within the inner tube remains fixed at 0.8 g/cm3. The reactivity reduces as water is added to this region, indicating that the most reactive condition is with void between the tubes.
In Series 3, the water density both inside and between the tubes is assumed to be exactly the same and varied between 0 and 1.0 g/cm3. These cases are less reactive than Case F9 in Series 1.
In Series 4, the moderation conditions of Series 1 are repeated except with the insulation modeled. The maximum reactivity occurs for Case J7 for a water density of 0.6 g/cm3, with a maximum ks = 0.73476. This case is slightly more reactive than Case F9, in which no insulation was modeled.
In Series 5, the most reactive case from Series 4 (Case J7) is modified so that the water density between the insulation and the outer tube is varied between 0 and 1.0 g/cm3, while the water density within the inner tube remains fixed at 0.6 g/cm3. The reactivity decreases as water is added to this region.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 In Series 6, the moderation conditions of Series 1 are repeated except with the FHE neoprene modeled. It was determined in the HAC single package analysis that neoprene will lower the reactivity due to absorption in the chlorine. Therefore, the neoprene is conservatively modeled without chlorine. The maximum reactivity occurs for Case L8, with ks = 0.73297, an increase of 3.6 mk when compared to Case F9. This increase is only slightly above statistical fluctuation, so it may be concluded that the presence of neoprene has at most a small effect on the reactivity. A water density of 0.7 g/cm3 within the inner tube is utilized in the most reactive case. No cases are performed with the neoprene homogeneously mixed into the water because this scenario is already implicitly considered using the variable water density search within the inner tube.
In Series 7, insulation and neoprene are combined in the same model with variable density water inside the inner pipe, as the presence of both insulation and neoprene slightly increased the reactivity when treated separately. The maximum reactivity occurs for Case M8, with ks =
0.73599. This case is slightly more reactive than the cases in which insulation and neoprene are addressed separately.
In Series 8, for completeness, void is modeled in the inner tube, while the water density is allowed to vary between the tubes. Chlorine-free neoprene is utilized to increase moderation in the inner tube, but insulation is ignored to maximize the amount of water between the tubes. The peak reactivity for Series 8 is the lowest of all nine series of calculations.
In Series 9, Series 7 is repeated with the channel width increased to 0.089-in. A channel width of 0.089-in is the maximum local channel width (0.087-in) with an additional margin of 0.002-in.
The larger channel width is achieved by reducing the cladding thickness. Case O5 is the most reactive of all computational series.
In conclusion, Case O5 is the most reactive, with ks = 0.74531. This result is below the USL of 0.9209. Case O5 has fully moderated fuel elements, 0.7 g/cm3 water in the inner tube, insulation and chlorine-free neoprene, void between the insulation and outer tube, fuel elements that have been rotated and moved to the center of the array, and a channel width of 0.089-in. Note that this result is lower than the maximum NCT array case because the HAC and NCT array models are quite similar, except the NCT array uses a much larger 9x9x1 configuration.
6.6.1.2 Loose Plate Basket Payload It was established in the criticality analysis for the ATR fuel element that the NCT array calculations bound the HAC array calculations. This result is obtained because a 9x9x1 array is utilized in the NCT calculations, while a smaller 5x5x1 array is utilized in the HAC array calculations. Water moderation is modeled in both the NCT and HAC array calculations within the inner tube, although additional moderation is allowed in the HAC cases between the inner and outer tubes. Therefore, the HAC array calculations are performed only for completeness, as these calculations will not be bounding.
In all the HAC array models, the loose plate basket is filled with full-density water, as it has been established in the NCT array analysis that full-density water moderation within the basket maximizes the reactivity. The internal plate arrangement determined in the NCT array calculations to be the most reactive (Case LC21 for plate type 5) is used in all HAC array models. Also, the loose plate basket is modeled pushed to the center of the array to maximize reactivity, as shown in Figure 6.6-2.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 Four series of calculations are performed that utilize different moderations conditions. Results for all cases are provided in Table 6.6-2.
Series 1: In Series 1, the insulation is modeled, and void is modeled between the insulation and the outer tube. The water density between the basket and the inner tube is varied between 0 and 1.0 g/cm3. The maximum reactivity is achieved for Case LK9, with ks = 0.69792. The water density for this case is 0.8 g/cm3.
Series 2: In Series 2, the most reactive case from Series 1 (Case LK9) is run with variable density water between the insulation and the outer tube. The reactivity decreases when water is added to this region.
Series 3: In Series 3, Series 1 is repeated, except the insulation is replaced with void. The maximum reactivity is close to but bounded by the maximum reactivity from Series 1.
Series 4: In Series 4, the insulation is not modeled, and the same water density is modeled both between the inner and outer tubes, and between the basket and inner tube. The maximum reactivity is significantly less than the maximum reactivity from Series 1.
In conclusion, the maximum reactivity is from Case LK9, with ks = 0.69792, in which full-density water is modeled within the basket, 0.8 g/cm3 water is modeled between the basket and the inner tube, and void between the insulation and the outer tube. This value is less than the USL of 0.9209.
6.6.2 HAC Array Results Following are the tabulated results for the HAC array cases. The most reactive configuration in each series is listed in boldface.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                Rev. 16, May 2021 Table 6.6 HAC Array Results, Fuel Element Water        Water Density Water Density      Density          Between Case                    Between      Inside Inner          Plates                            ks ID    Filename      Tubes (g/cm3) Tube (g/cm3)            (g/cm3)        keff            (k+2)
Series 1: Variable water density in inner tube (no insulation, no neoprene)
F1    HA_S0P000            0                0                1.0      0.57908    0.00102  0.58112 F2    HA_S0P010            0              0.1              1.0      0.63182    0.00112  0.63406 F3    HA_S0P020            0              0.2              1.0      0.66922    0.00124  0.67170 F4    HA_S0P030            0              0.3              1.0      0.69357    0.00121  0.69599 F5    HA_S0P040            0              0.4              1.0      0.71180    0.00116  0.71412 F6    HA_S0P050            0              0.5              1.0      0.72106    0.00120  0.72346 F7    HA_S0P060            0              0.6              1.0      0.72553    0.00122  0.72797 F8    HA_S0P070            0              0.7              1.0      0.72706    0.00112  0.72930 F9    HA_S0P080            0              0.8              1.0      0.72695    0.00119  0.72933 F10  HA_S0P090            0              0.9              1.0      0.72116    0.00110  0.72336 F11  HA_S0P100            0              1.0              1.0      0.71826    0.00123  0.72072 Series 2: Case F9 with variable density water between tubes F9    HA_S0P080            0              0.8              1.0      0.72695    0.00119  0.72933 G1    HA_P80S010          0.1              0.8              1.0      0.70205    0.00112  0.70429 G2    HA_P80S020          0.2              0.8              1.0      0.67677    0.00125  0.67927 G3    HA_P80S030          0.3              0.8              1.0      0.65374    0.00113  0.65600 G4    HA_P80S040          0.4              0.8              1.0      0.63121    0.00114  0.63349 G5    HA_P80S050          0.5              0.8              1.0      0.60791    0.00104  0.60999 G6    HA_P80S060          0.6              0.8              1.0      0.59303    0.00111  0.59525 G7    HA_P80S070          0.7              0.8              1.0      0.57461    0.00109  0.57679 G8    HA_P80S080          0.8              0.8              1.0      0.56082    0.00110  0.56302 G9    HA_P80S090          0.9              0.8              1.0      0.54767    0.00102  0.54971 G10    HA_P80S100          1.0              0.8              1.0      0.53613    0.00108  0.53829 Series 3: Matching water density inside and between tubes F1    HA_S0P000            0                0                1.0      0.57908    0.00102  0.58112 H1    HA_SP010            0.1              0.1              1.0      0.64719    0.00115  0.64949 H2    HA_SP020            0.2              0.2              1.0      0.66047    0.00115  0.66277 H3    HA_SP030            0.3              0.3              1.0      0.64457    0.00112  0.64681 H4    HA_SP040            0.4              0.4              1.0      0.62648    0.00117  0.62882 H5    HA_SP050            0.5              0.5              1.0      0.60286    0.00112  0.60510 H6    HA_SP060            0.6              0.6              1.0      0.58814    0.00116  0.59046 H7    HA_SP070            0.7              0.7              1.0      0.57337    0.00106  0.57549 H8    HA_SP080            0.8              0.8              1.0      0.56082    0.00110  0.56302 H9    HA_SP090            0.9              0.9              1.0      0.55245    0.00122  0.55489 H10    HA_SP100            1.0              1.0              1.0      0.54360    0.00100  0.54560 (continued) 6-60
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                              Rev. 16, May 2021 Table 6.6 HAC Array Results, Fuel Element (continued)
Water      Water Density Water Density      Density        Between Case                    Between        Inside Inner      Plates                            ks ID    Filename      Tubes (g/cm3) Tube (g/cm3)          (g/cm3)        keff            (k+2)
Series 4: Repeat of Series 1 with insulation J1  HA_DS0P000            0                0              1.0      0.58824  0.00116  0.59056 J2  HA_DS0P010            0              0.1            1.0      0.63716  0.00111  0.63938 J3  HA_DS0P020            0              0.2            1.0      0.67403  0.00118  0.67639 J4  HA_DS0P030            0              0.3            1.0      0.69920  0.00130  0.70180 J5  HA_DS0P040            0              0.4            1.0      0.71665  0.00116  0.71897 J6  HA_DS0P050            0              0.5            1.0      0.72388  0.00117  0.72622 J7    HA_DS0P060            0              0.6            1.0      0.73230  0.00123  0.73476 J8  HA_DS0P070            0              0.7            1.0      0.73178  0.00112  0.73402 J9  HA_DS0P080            0              0.8            1.0      0.72965  0.00124  0.73213 J10  HA_DS0P090            0              0.9            1.0      0.72638  0.00107  0.72852 J11  HA_DS0P100            0              1.0            1.0      0.71985  0.00113  0.72211 Series 5: Case J7 with variable density water between insulation and outer tube J7    HA_DS0P090            0              0.6            1.0      0.73230  0.00123  0.73476 K1    HA_DP60S010        0.1              0.6            1.0      0.72284  0.00123  0.72530 K2    HA_DP60S020        0.2              0.6            1.0      0.71587  0.00120  0.71827 K3    HA_DP60S030        0.3              0.6            1.0      0.71029  0.00118  0.71265 K4    HA_DP60S040        0.4              0.6            1.0      0.70002  0.00117  0.70236 K5    HA_DP60S050        0.5              0.6            1.0      0.69370  0.00122  0.69614 K6    HA_DP60S060        0.6              0.6            1.0      0.68266  0.00111  0.68488 K7    HA_DP60S070        0.7              0.6            1.0      0.67122  0.00112  0.67346 K8    HA_DP60S080        0.8              0.6            1.0      0.66359  0.00115  0.66589 K9    HA_DP60S090        0.9              0.6            1.0      0.65393  0.00111  0.65615 K10    HA_DP60S100        1.0              0.6            1.0      0.64595  0.00116  0.64827 Series 6: Repeat of Series 1 with neoprene L1    HA_N2S0P000          0                0              1.0      0.60058  0.00113  0.60284 L2    HA_N2S0P010          0              0.1            1.0      0.64323  0.00119  0.64561 L3    HA_N2S0P020          0              0.2            1.0      0.68153  0.00118  0.68389 L4    HA_N2S0P030          0              0.3            1.0      0.70640  0.00120  0.70880 L5    HA_N2S0P040          0              0.4            1.0      0.71669  0.00124  0.71917 L6    HA_N2S0P050          0              0.5            1.0      0.72733  0.00117  0.72967 L7    HA_N2S0P060          0              0.6            1.0      0.72872  0.00122  0.73116 L8    HA_N2S0P070          0              0.7            1.0      0.73069  0.00114  0.73297 L9    HA_N2S0P080          0              0.8            1.0      0.73081  0.00107  0.73295 L10  HA_N2S0P090          0              0.9            1.0      0.72692  0.00129  0.72950 L11  HA_N2S0P100          0              1.0            1.0      0.72371  0.00122  0.72615 (continued) 6-61
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                              Rev. 16, May 2021 Table 6.6 HAC Array Results, Fuel Element (concluded)
Water        Water        Water Density      Density      Density Between        Inside      Between Case                        Tubes    Inner Tube      Plates                            ks ID      Filename          (g/cm3)      (g/cm3)      (g/cm3)      keff            (k+2)
Series 7: Repeat of Series 1 with insulation and neoprene M1    HA_DNS0P000            0            0            1.0      0.60377    0.00107  0.60591 M2    HA_DNS0P010            0            0.1          1.0      0.64940    0.00107  0.65154 M3    HA_DNS0P020            0            0.2          1.0      0.68596    0.00110  0.68816 M4    HA_DNS0P030            0            0.3          1.0      0.70846    0.00115  0.71076 M5    HA_DNS0P040            0            0.4          1.0      0.72168    0.00122  0.72412 M6    HA_DNS0P050            0            0.5          1.0      0.73000    0.00124  0.73248 M7    HA_DNS0P060            0            0.6          1.0      0.73182    0.00122  0.73426 M8    HA_DNS0P070            0            0.7          1.0      0.73365    0.00117  0.73599 M9    HA_DNS0P080            0            0.8          1.0      0.73187    0.00127  0.73441 M10    HA_DNS0P090            0            0.9          1.0      0.73006    0.00112  0.73230 M11    HA_DNS0P100            0            1.0          1.0      0.72332    0.00122  0.72576 Series 8: Case L1 with variable density water between tubes L1    HA_N2S0P000            0            0            1.0      0.60058    0.00113  0.60284 N1    HA_N2P0S010          0.1            0            1.0      0.63054    0.00107  0.63268 N2    HA_N2P0S020          0.2            0            1.0      0.62961    0.00118  0.63197 N3    HA_N2P0S030          0.3            0            1.0      0.61939    0.00113  0.62165 N4    HA_N2P0S040          0.4            0            1.0      0.60776    0.00108  0.60992 N5    HA_N2P0S050          0.5            0            1.0      0.58874    0.00108  0.59090 N6    HA_N2P0S060          0.6            0            1.0      0.57308    0.00109  0.57526 N7    HA_N2P0S070          0.7            0            1.0      0.55837    0.00107  0.56051 N8    HA_N2P0S080          0.8            0            1.0      0.54139    0.00101  0.54341 N9    HA_N2P0S090          0.9            0            1.0      0.52714    0.00106  0.52926 N10  HA_N2P0S100          1.0            0            1.0      0.51600    0.00114  0.51828 Series 9: Repeat of Series 7 with a channel width of 0.089-in O1  HA_DNS0P030_C89        0            0.3          1.0      0.71885    0.00118  0.72121 O2  HA_DNS0P040_C89        0            0.4          1.0      0.73301    0.00114  0.73529 O3  HA_DNS0P050_C89        0            0.5          1.0      0.73889    0.00124  0.74137 O4  HA_DNS0P060_C89        0            0.6          1.0      0.74172    0.00121  0.74414 O5  HA_DNS0P070_C89        0            0.7          1.0      0.74299    0.00116  0.74531 O6  HA_DNS0P080_C89        0            0.8          1.0      0.74192    0.00125  0.74442 O7  HA_DNS0P090_C89        0            0.9          1.0      0.73527    0.00105  0.73737 O8  HA_DNS0P100_C89        0            1.0          1.0      0.73282    0.00110  0.73502 6-62
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                Rev. 16, May 2021 Table 6.6 HAC Array Results, Plate 5 Water          Water          Water Density      Density        Density Between      Inside Inner      Between Case                        Tubes          Tube            Plates                            ks ID    Filename          (g/cm3)        (g/cm3)        (g/cm3)        keff          (k+2)
Series 1: Variable water density in inner tube, with insulation LK1  HA_N5DS0P000          0              0              1.0      0.53784  0.00096  0.53976 LK2  HA_N5DS0P010          0            0.1              1.0      0.58878  0.00098  0.59074 LK3  HA_N5DS0P020          0            0.2              1.0      0.62946  0.00101  0.63148 LK4  HA_N5DS0P030          0            0.3              1.0      0.65858  0.00102  0.66062 LK5  HA_N5DS0P040          0            0.4              1.0      0.67685  0.00100  0.67885 LK6  HA_N5DS0P050          0            0.5              1.0      0.68901  0.00102  0.69105 LK7  HA_N5DS0P060          0            0.6              1.0      0.69483  0.00107  0.69697 LK8  HA_N5DS0P070          0            0.7              1.0      0.69266  0.00120  0.69506 LK9  HA_N5DS0P080          0            0.8              1.0      0.69576  0.00108  0.69792 LK10  HA_N5DS0P090          0            0.9              1.0      0.69250  0.00105  0.69460 LK11  HA_N5DS0P100          0            1.0              1.0      0.68585  0.00104  0.68793 Series 2: Case LK9 with variable density water between tubes.
LK9  HA_N5DS0P080          0            0.8              1.0      0.69576  0.00108  0.69792 LM1  HA_N5DP80S010          0.1            0.8              1.0      0.68989  0.00106  0.69201 LM2  HA_N5DP80S020          0.2            0.8              1.0      0.67989  0.00107  0.68203 LM3  HA_N5DP80S030          0.3            0.8              1.0      0.67352  0.00098  0.67548 LM4  HA_N5DP80S040          0.4            0.8              1.0      0.66658  0.00105  0.66868 LM5  HA_N5DP80S050          0.5            0.8              1.0      0.65700  0.00105  0.65910 LM6  HA_N5DP80S060          0.6            0.8              1.0      0.64893  0.00118  0.65129 LM7  HA_N5DP80S070          0.7            0.8              1.0      0.64141  0.00106  0.64353 LM8  HA_N5DP80S080          0.8            0.8              1.0      0.63415  0.00099  0.63613 LM9  HA_N5DP80S090          0.9            0.8              1.0      0.62748  0.00103  0.62954 LM10  HA_N5DP80S100          1.0            0.8              1.0      0.62100  0.00094  0.62288 Series 3: Repeat of Series 1, no insulation LN1    HA_N5S0P000          0              0              1.0      0.53334  0.00092  0.53518 LN2    HA_N5S0P010          0            0.1              1.0      0.58456  0.00091  0.58638 LN3    HA_N5S0P020          0            0.2              1.0      0.62421  0.00108  0.62637 LN4    HA_N5S0P030          0            0.3              1.0      0.65402  0.00109  0.65620 LN5    HA_N5S0P040          0            0.4              1.0      0.67129  0.00108  0.67345 LN6    HA_N5S0P050          0            0.5              1.0      0.68550  0.00108  0.68766 LN7    HA_N5S0P060          0            0.6              1.0      0.69042  0.00106  0.69254 LN8  HA_N5S0P070            0            0.7              1.0      0.69145  0.00104  0.69353 LN9    HA_N5S0P080          0            0.8              1.0      0.69071  0.00101  0.69273 LN10    HA_N5S0P090          0            0.9              1.0      0.68925  0.00102  0.69129 LN11    HA_N5S0P100          0            1.0              1.0      0.68493  0.00116  0.68725 (continued) 6-63
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                            Rev. 16, May 2021 Table 6.6 HAC Array Results, Plate 5 (concluded)
Water                        Water Density        Water        Density Between        Density      Between Case                      Packages    Inside Pipe      Plates                          ks ID    Filename          (g/cm3)      (g/cm3)        (g/cm3)      keff            (k+2)
Series 4: No insulation, matching water densities inside and between tubes LN1  HA_N5S0P000            0            0              1.0      0.53334  0.00092  0.53518 LO1    HA_N5SP010            0.1          0.1            1.0      0.59685  0.00097  0.59879 LO2    HA_N5SP020            0.2          0.2            1.0      0.61533  0.00096  0.61725 LO3    HA_N5SP030            0.3          0.3            1.0      0.60844  0.00108  0.61060 LO4    HA_N5SP040            0.4          0.4            1.0      0.59462  0.00099  0.59660 LO5    HA_N5SP050            0.5          0.5            1.0      0.57802  0.00107  0.58016 LO6    HA_N5SP060            0.6          0.6            1.0      0.56514  0.00107  0.56728 LO7    HA_N5SP070            0.7          0.7            1.0      0.55116  0.00106  0.55328 LO8    HA_N5SP080            0.8          0.8            1.0      0.54262  0.00093  0.54448 LO9    HA_N5SP090            0.9          0.9            1.0      0.53400  0.00102  0.53604 LO10    HA_N5SP100            1.0          1.0            1.0      0.52785  0.00106  0.52997 6-64
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                              Rev. 16, May 2021 Series 1: Array with variable density water in inner tube, and void between tubes. No insulation modeled.
Series 5: Array with 0.6 g/cm3 water in inner tube and variable density water between tubes.
Insulation is modeled.
Figure 6.6 HAC Array Geometry Examples, Fuel Element 6-65
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                          Rev. 16, May 2021 Series 1: Array with variable density water in inner tube, and void between tubes. Insulation is modeled.
Series 3: Array with variable density water in inner tube, and void between tubes. No insulation modeled.
Figure 6.6 HAC Array Geometry Examples, Loose Plate Basket 6-66
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 6.7 Fissile Material Packages for Air Transport The applicable licensing requirements for air transport of fissile material are contained in 10 CFR 71.55(f) and IAEA Safety Standard SSR-6, paragraph 683. For air transport, no structural integrity is credited for the hypothetical accident conditions. Rather, the fissile material from a single package is modeled as a sphere that is optimally moderated and reflected by the packaging materials. The sphere is reflected with 20 cm of water. Per the licensing requirements of 10 CFR 71.55(f) and IAEA Safety Standard SSR-6, paragraph 683, environmental water is not included in the fissile sphere.
The air transport analysis is intended to bound all current and projected future payloads. 2000 g U-235 in HEU is conservatively modeled. The non-uranium elements that may be present in the fuel matrix (e.g., aluminum, silicon, molybdenum), and structural materials of the fuel elements and package (e.g., aluminum, steel) are conservatively neglected within the fissile mixture, which minimizes parasitic neutron absorption. Omitting these materials from the fissile sphere also minimizes the size of the fissile sphere, which minimizes neutron leakage. Both of these effects conservatively maximize the reactivity. The U-235 weight percent is modeled at the maximum value of 94%. Consistent with the ATR fuel element analysis, representative weight percents of 0.6% and 0.35% are utilized for U-234 and U-236, respectively, and the balance (5.05%) is modeled as U-238.
The package contains up to 100 g polyethylene (CH2), and the sum of neoprene (C4H5Cl) and cellulosic material (C6H10O5) (such as kraft paper and cardboard) is limited to 4000 g. Because polyethylene, neoprene, and cellulosic material contain hydrogen, these materials act as a moderator and are explicitly addressed in the analysis.
The fissile sphere is modeled as a mixture of uranium, polyethylene, neoprene and/or cellulosic material. For an air transport analysis, environmental water is not included. The total mass of uranium based on 94% enrichment is 2000 g U-235 / 0.94 = 2127.7 g uranium. The theoretical material densities are 19.0 g/cm3 uranium metal, 0.92 g/cm3 polyethylene, 1.23 g/cm3 neoprene, and 0.44 g/cm3 for cellulosic material (density of kraft paper). Based upon the mass inputs and the material densities, the volume of uranium, polyethylene, neoprene, and cellulosic material are summed to create the total volume for the fissile sphere, which is used to compute atom densities for each isotope.
The atom densities for several different example mixtures are provided in Table 6.7-1. The total volume V is computed as V = VU + Vpoly + Vcellulosic + Vneoprene, and the radius R of the fissile material is then:
3 4
The gram density for each material within the sphere is computed as the mass of the material divided by the total volume (V), which is then used to compute the atom density of each material. These atom densities are then summed to create the overall mixture atom densities presented in Table 6.7-1.
In the first series of cases, the uranium is homogeneously mixed with various quantities of polyethylene, neoprene, and cellulosic material in a single fissile sphere reflected with 20 cm of 6-67
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 water, see Figure 6.7-1. These results are summarized in Table 6.7-2. Case AIR01 is pure uranium with no hydrogenous material, with ks = 0.51683. Cases AIR02 through AIR06 include increasing amounts of polyethylene (no neoprene or cellulosic material) up to the maximum value of 100 g polyethylene. The reactivity decreases compared to Case AIR01 when < 80 g polyethylene is added, but reactivity is maximized with the full 100 g polyethylene, with ks =
0.52233.
Cases AIR07 through AIR14 include increasing amounts of neoprene (no polyethylene or cellulosic material) up to the maximum value of 4000 g neoprene. The reactivity decreases compared to Case AIR01 when < 1500 g of neoprene is added, but reactivity is maximized with the full 4000 g neoprene, with ks = 0.57035.
Cases AIR15 through AIR22 include increasing amounts of cellulosic material (no polyethylene or neoprene) up to the maximum value of 4000 g cellulosic material. The reactivity decreases compared to Case AIR01 when < 3000 g of cellulosic material is added, but reactivity is maximized with the full 4000 g cellulosic material, with ks = 0.55065.
For all three hydrogenous materials, reactivity initially decreases for smaller amounts of moderating material. The most likely reason that reactivity initially decreases when small amounts of hydrogenous material are added is that adding moderator greatly enlarges the volume of the fissile sphere. For small amounts of moderator, the increase in neutron leakage due to the larger sphere results in a net reduction in reactivity despite the increase in moderation. For larger quantities of moderator, the enhanced moderation increases the reactivity despite the increased leakage.
The sum of neoprene and cellulosic material is limited to 4000 g. Comparing Cases AIR14 and AIR22, it is concluded that neoprene is a superior moderator than cellulosic material. Therefore, it is conservative to model the 4000 g of neoprene/cellulosic material as 4000 g neoprene. In Case AIR23, both 100 g polyethylene and 4000 g neoprene are included in the fissile sphere, with ks = 0.58222. This is the most reactive single fissile sphere case.
The above models do not include the packaging structural materials, which may also act as a reflector. The maximum weight of a loaded package is 290 lbs (see Section 1.2.1.2). This weight includes the fuel, stainless steel structural members, aluminum fuel support structures, insulation, etc. Therefore, 300 lbs (136 kg) bounds the total mass of stainless steel in the package by a large margin and is used as a reflector outside the fissile sphere. The stainless steel density is 7.94 g/cm3, resulting in a steel reflector thickness of 7.61 cm (see Figure 6.7-2) if added to the most reactive single fissile sphere model (Case AIR23). A 20 cm water reflector is modeled outside the steel reflector. When the steel reflector is added (Case AIR24), ks =
0.55311, which is less reactive than without the steel reflector. Therefore, the reflection provided by the structural materials of the packaging may be neglected.
The single fissile sphere is undermoderated with 4000 g neoprene and 100 g polyethylene.
Moderation may be further enhanced by modeling the fissile material in two regions, an inner sphere and outer shell, with all of the moderating material in the inner sphere (see Figure 6.7-3).
The inner sphere contains a mass M grams U-235, 4000 g neoprene, and 100 g polyethylene.
The outer shell contains (2000 - M) grams U-235 with no moderating material. Therefore, the system always contains 2000 g U-235. The inner sphere in the two region model achieves greater moderation than the single sphere model because less U-235 is moderated with the full mass of moderating material.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                              Rev. 16, May 2021 The atom densities are computed using the same method developed for the single region model, although the mass of U-235 is now varied. Sample atom densities for the moderated inner sphere for several different U-235 masses are summarized in Table 6.7-3, as well as the outer shell composition.
The two-region model results are summarized in Table 6.7-4. The two-region model achieves a higher reactivity than the single-region model. The reactivity is maximized with 1500 g U-235 in the inner sphere and 500 g U-235 in the outer shell (Case AIR40), with ks = 0.60739. The most reactive air transport case is far below the USL.
Table 6.7 Example Atom Densities, Single Fissile Sphere 2000 g U-235          2000 g U-235        2000 g U-235        2000 g U-235 0 g polyethylene      100 g polyethylene  0 g polyethylene    0 g polyethylene 0 g neoprene          0 g neoprene    4000 g neoprene        0 g neoprene 0 g cellulosic        0 g cellulosic      0 g cellulosic    4000 g cellulosic Isotope          (atom/b-cm)            (atom/b-cm)        (atom/b-cm)        (atom/b-cm)
U-234          2.9333E-04            1.4885E-04          9.7644E-06          3.5650E-06 U-235          4.5759E-02            2.3220E-02          1.5232E-03          5.5614E-04 U-236          1.6965E-04            8.6091E-05          5.6475E-06          2.0619E-06 U-238          2.4273E-03            1.2317E-03          8.0799E-05          2.9500E-05 H                  -                3.8910E-02          4.0439E-02          1.6124E-02 C                  -                1.9455E-02          3.2351E-02          9.6743E-03 O                  -                      -                  -              8.0619E-03 Cl                  -                      -              8.0877E-03              -
Total          4.8649E-02            8.3052E-02          8.2497E-02          3.4451E-02 6-69
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                                      Rev. 16, May 2021 Table 6.7 Air Transport Results, Single Fissile Sphere Poly-Radius ethylene    Neoprene  Cellulosic                      ks Case ID        Filename              (cm)  Mass (g)      Mass (g) Mass (g)      k            (k+2)
AIR01    AIR_P0_N0_K0              2.9901    0              0      0      0.51621 0.00031    0.51683 AIR02    AIR_P020                  3.1723    20              0      0      0.50988 0.00031    0.51050 AIR03    AIR_P040                  3.3356    40              0      0      0.50946 0.00032    0.51010 AIR04    AIR_P060                  3.4844    60              0      0      0.51250 0.00033    0.51316 AIR05    AIR_P080                  3.6214    80              0      0      0.51649 0.00033    0.51715 AIR06    AIR_P100                  3.7488    100            0      0      0.52167 0.00033    0.52233 AIR07    AIR_N0500                4.9837    0            500      0      0.48059 0.00034    0.48127 AIR08    AIR_N1000                6.0443    0            1000      0      0.50801 0.00035    0.50871 AIR09    AIR_N1500                6.8247    0            1500      0      0.52854 0.00036    0.52926 AIR10    AIR_N2000                7.4585    0            2000      0      0.54342 0.00038    0.54418 AIR11    AIR_N2500                7.9998    0            2500      0      0.55460 0.00036    0.55532 AIR12    AIR_N3000                8.4763    0            3000      0      0.56208 0.00036    0.56280 AIR13    AIR_N3500                8.9046    0            3500      0      0.56762 0.00035    0.56832 AIR14    AIR_N4000                9.2952    0            4000      0      0.56965 0.00035    0.57035 Note: All models reflected with 20 cm water.
(continued) 6-70
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                                                                Rev. 16, May 2021 Table 6.7 Air Transport Results, Single Fissile Sphere (concluded)
Poly-Radius        ethylene      Neoprene        Cellulosic                                ks Case ID        Filename              (cm)        Mass (g)      Mass (g)        Mass (g)        k                  (k+2)
AIR15    AIR_K0500                  6.6820            0              0              500        0.43112      0.00034    0.43180 AIR16    AIR_K1000                  8.2912            0              0            1000        0.45924      0.00036    0.45996 AIR17    AIR_K1500                  9.4413            0              0            1500        0.48136      0.00036    0.48208 AIR18    AIR_K2000                10.3639            0              0            2000        0.49829      0.00037    0.49903 AIR19    AIR_K2500                11.1463            0              0            2500        0.51439      0.00038    0.51515 AIR20    AIR_K3000                11.8319            0              0            3000        0.52755      0.00039    0.52833 AIR21    AIR_K3500                12.4462            0              0            3500        0.53825        0.0004    0.53905 AIR22    AIR_K4000                13.0052            0              0            4000        0.54989      0.00038    0.55065 AIR23    AIR_NP                    9.3942          100          4000              0        0.58148      0.00037    0.58222 Case AIR24 features a steel reflector between the fissile sphere and outer water reflector AIR24    AIR_NP_RSS                9.3942          100          4000              0        0.55239      0.00036    0.55311 Note: All models reflected with 20 cm water.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                              Rev. 16, May 2021 Table 6.7 Example Atom Densities, Two-Region Model Inner Region                        Outer Region 500 g U-235        1000 g U-235      1500 g U-235 100 g polyethylene 100 g polyethylene 100 g polyethylene 4000 g neoprene    4000 g neoprene    4000 g neoprene Isotope        (atom/b-cm)        (atom/b-cm)        (atom/b-cm)    Uranium Metal U-234          2.4233E-06        4.8069E-06        7.1517E-06      2.9333E-04 U-235          3.7803E-04        7.4986E-04        1.1157E-03      4.5759E-02 U-236          1.4016E-06        2.7802E-06        4.1364E-06      1.6965E-04 U-238          2.0052E-05        3.9776E-05        5.9180E-05      2.4273E-03 H            4.2678E-02        4.2328E-02        4.1984E-02            -
C            3.3382E-02        3.3108E-02        3.2839E-02            -
Cl            8.0288E-03        7.9630E-03        7.8983E-03            -
Total          8.4490E-02        8.4197E-02        8.3908E-02      4.8649E-02 6-72
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                                                                  Rev. 16, May 2021 Table 6.7 Air Transport Results, Two-Region Model Radius        U-235 Mass      U-235 Mass                                  ks Case ID          Filename          Inner (cm)          Inner (g)      Outer (g)    k                        (k+2)
AIR30        AIR_MI0500                9.3179            500            1500    0.55303        0.00035      0.55373 AIR31        AIR_MI0600                9.3230            600            1400    0.56436        0.00035      0.56506 AIR32        AIR_MI0700                9.3281            700            1300    0.57413        0.00035      0.57483 AIR33        AIR_MI0800                9.3332            800            1200    0.58213        0.00037      0.58287 AIR34        AIR_MI0900                9.3383            900            1100    0.58823        0.00036      0.58895 AIR35        AIR_MI1000                9.3434            1000            1000    0.59313        0.00036      0.59385 AIR36        AIR_MI1100                9.3485            1100            900    0.59856        0.00037      0.59930 AIR37        AIR_MI1200                9.3536            1200            800    0.60211        0.00038      0.60287 AIR38        AIR_MI1300                9.3587            1300            700    0.60424        0.00038      0.60500 AIR39        AIR_MI1400                9.3638            1400            600    0.60641        0.00038      0.60717 AIR40        AIR_MI1500                9.3689            1500            500    0.60663        0.00038      0.60739 AIR41        AIR_MI1600                9.3740            1600            400    0.60610        0.00035      0.60680 AIR42        AIR_MI1700                9.3790            1700            300    0.60433        0.00037      0.60507 AIR43        AIR_MI1800                9.3841            1800            200    0.60075        0.00039      0.60153 AIR44        AIR_MI1900                9.3892            1900            100    0.59442        0.00039      0.59520 AIR45        AIR_MI1995                9.3940            1995              5      0.58274        0.00036      0.58346 Notes:
(1) The total U-235 mass is 2000 g in all models.
(2) All models contain 4000 g neoprene and 100 g polyethylene in the inner sphere. The outer shell is uranium metal.
(3) The outer radius of the outer fissile shell is 9.3942 cm in all models.
(4) All cases reflected with 20 cm water.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                              Rev. 16, May 2021 Water Reflector Moderated Fuel Region Figure 6.7 Air Transport Model, Single Region Model 6-74
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                            Rev. 16, May 2021 Water Reflector Steel Reflector Moderated Fuel Region Figure 6.7 Air Transport Model with Steel Reflector 6-75
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                Rev. 16, May 2021 Water Reflector Moderated Fuel Region Uranium Metal Shell Note: This figure is not to scale. The thickness of the uranium metal shell has been exaggerated for illustrative purposes.
Figure 6.7 Air Transport Model, Two-Region Model 6-76
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                        Rev. 16, May 2021 6.8 Benchmark Evaluations The MCNP, Version 5, Monte Carlo computer code8 with point-wise ENDF/B-V, -VI, and -VII cross sections has been used extensively in criticality evaluations. The uranium isotopes utilize preliminary ENDF/B-VII cross section data that are considered by Los Alamos National Laboratory to be more accurate than ENDF/B-VI cross sections. ENDF/B-V cross sections are utilized for chromium, nickel, and iron because natural composition ENDF/B-VI cross sections are not available for these elements. The remaining isotopes utilize ENDF/B-VI cross sections.
This section justifies the validity of this computation tool and data library combination for application to the ATR FFSC criticality analysis and a bias factor is obtained from these calculations of the critical experiments.
The MCNP code uses room temperature continuous-energy (point-wise) cross sections that are thoroughly documented in Appendix G of the manual. These cross sections are defined with a high-energy resolution that describes each resolved cross section resonance for the isotope. All of the cross-sections used for these analyses were generated from the U.S. Evaluated Nuclear Data Files (ENDF/B).
The validation of the point-wise cross sections is conducted using 35 experimental criticality benchmarks applicable to ATR plate fuels transported in the ATR FFSC. These experiments are supplemented with 21 fast and intermediate spectrum experiments to bound the parameters of the air transport analysis. The statistical analysis of the benchmark experiments results in a USL of 0.9209.
6.8.1 Applicability of Benchmark Experiments Plate Fuel Analysis The experimental benchmarks are summarized in the OECD Nuclear Energy Agencys International Handbook of Evaluated Criticality Safety Benchmark Experiments9. Each experiment is discussed in detail in the Handbook. It includes estimates of the uncertainty in the measurements, detailed information regarding dimensions and material compositions, comparisons between the multiplication factor calculated by various computer codes, and a list of input files that were used in their calculations.
The critical experiment benchmarks are selected based upon their similarity to the ATR FFSC and contents. The important selection parameters are high-enriched uranium plate-type fuel with a thermal spectrum. Thirty-five (35) benchmarks that meet these criteria are selected from the Handbook. The titles for all utilized experiments are listed in Table 6.8-1. Note that the benchmark from HEU-MET-THERM-022 is for the Advanced Test Reactor itself, so the fuel configuration in this benchmark is essentially the same as the fuel modeled in the packaging analysis.
Ideally, benchmarks would be limited to those with a fuel matrix of UAlx and aluminum, aluminum cladding, and no absorbers, consistent with the ATR criticality models. Experiment 8
MCNP5, MCNP - A General Monte Carlo N-Particle Transport Code, Version 5; Volume II: Users Guide, LA-CP-03-0245, Los Alamos National Laboratory, April, 2003.
9 OECD Nuclear Energy Agency, International Handbook of Evaluated Criticality Safety Benchmark Experiments, NEA/NSC/DOC(95)03, September, 2006.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                        Rev. 16, May 2021 set HEU-MET-THERM-006 consists of 23 benchmark experiments. The first 16 experiments are directly applicable, although experiments 17 and 18 utilize thin cadmium sheets, and experiments 19 through 23 utilize uranium in solution in addition to the fuel plates. Experiment set HEU-COMP-THERM-022 consists of 11 benchmark experiments that utilize UO2 powder sintered with stainless steel, and stainless steel cladding. Experiments 1 through 5 do not utilize control rods, while experiments 6 through 11 utilize boron control rods. HEU-MET-THERM-022 is a detailed model of the ATR core using explicit ATR fuel elements very similar to the ATR fuel element model utilized in the criticality analysis. However, this full-core model necessarily contains absorber materials. Despite the presence of absorbers, because this benchmark utilizes ATR fuel, it is considered directly applicable to the ATR criticality analysis.
Therefore, of these 35 benchmarks, 17 benchmarks are directly applicable, while 18 benchmarks are applicable to a lesser degree. To compensate for the benchmarks that are not directly applicable, trending will be performed both on all 35 benchmark experiments (Set 1) and on the subset of 17 directly applicable benchmark experiments (Set 2). The USL selected is the minimum of both experimental sets.
Benchmark input files are either obtained from the Handbook or directly from Idaho National Laboratory (INL). The only changes made to the input files involve changing to a consistent set of cross section libraries, as needed. Review of the input files indicates that standard MCNP modeling techniques are employed. All but one of the input files consists of simple flat plates in various arrangements. The only benchmark that deviates from simple flat plates is the Advanced Test Reactor full-core model, which is directly applicable to the current analysis. These benchmark input files were developed by INL and have been used extensively for their internal criticality evaluations and are considered to be acceptable. Because the geometry and materials are modeled explicitly, any analyst properly modeling the experimental configuration in MCNP5 would obtain the same result within statistical fluctuation.
Air Transport Analysis A limited quantity of moderating material is used in the air transport analysis. As a result, the air transport models are closer to an intermediate or fast system and are generally outside the bounds of the plate fuel benchmark experiments described in the preceding paragraphs. Therefore, an additional 21 fast and intermediate benchmark experiments are used to supplement the plate-fuel benchmark experiments. The fast and intermediate benchmark experiments are primarily HEU, consistent with the air transport models.
The fast/intermediate spectrum experiments are summarized in Table 6.8-1. These fast/intermediate spectrum experiments are based upon the OECD Nuclear Energy Agencys International Handbook of Evaluated Criticality Safety Benchmark Experiments.
6.8.2 Bias Determination for Plate Fuels The USL is calculated by application of the USLSTATS computer program10. USLSTATS receives as input the keff as calculated by MCNP, the total 1- uncertainty (combined benchmark and MCNP uncertainties), and a trending parameter. Five trending parameters have been selected: (1) Energy of the Average neutron Lethargy causing Fission (EALF), (2) U-235 10 USLSTATS, USLSTATS: A Utility To Calculate Upper Subcritical Limits For Criticality Safety Applications, Version 1.4.2, Oak Ridge National Laboratory, April 23, 2003.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                Rev. 16, May 2021 number density, (3) channel width, (4) ratio of the number of hydrogen atoms in a unit cell to the number of U-235 atoms in a unit cell (H/U-235), and (5) plate pitch.
The uncertainty value, total, assigned to each case is a combination of the benchmark uncertainty for each experiment, bench, and the Monte Carlo uncertainty associated with the particular computational evaluation of the case, MCNP, or:
total = (bench2 + MCNP2)1/2 These values are input into the USLSTATS program in addition to the following parameters, which are the values recommended by the USLSTATS users manual:
P, proportion of population falling above lower tolerance level = 0.995 (note that this parameter is required input but is not utilized in the calculation of USL Method 1) 1-, confidence on fit = 0.95
  , confidence on proportion P = 0.95 (note that this parameter is required input but is not utilized in the calculation of USL Method 1) km, administrative margin used to ensure subcriticality = 0.05.
These data are followed by triplets of trending parameter value, computed keff, and uncertainty for each case. A confidence band analysis is performed on the data for each trending parameter using USL Method 1. The USL generated for each of the trending parameters utilized is provided in Table 6.8-2. All benchmark data used as input to USLSTATS are reported in Table 6.8-3.
In the following sections, the minimum USL computed for each parameter is identified, and the range of applicability is compared to the fuel element and loose plate models.
6.8.2.1 Energy of the Average neutron Lethargy causing Fission (EALF)
The EALF is used as the first trending parameter for the benchmark cases. The EALF comparison provides a means to observe neutron spectral dependencies or trends. The data for all 35 experiments of Set 1 are plotted in Figure 6.8-1. Over the range of applicability, the minimum USL is 0.9254 for Set 1 and 0.9212 for Set 2.
Range of Applicability, Fuel element models: All of the single package models and most of the NCT and HAC array models fall within the range of the applicability. The EALF of the most reactive fuel element model (Case E23) has an EALF of 1.39E-07 MeV, which is within the range of applicability. Models with significantly more void spaces or low water densities sometimes exceed the range of applicability (maximum EALF = 2.73E-07 MeV for Case E1),
although these cases are not the most reactive. Therefore, the EALF of the most reactive models is acceptably within the range of applicability of the benchmarks.
Range of Applicability, Loose plate models: The loose plate analysis is highly moderated, and the EALF of the models fall within the range of applicability of the benchmark experiments with few exceptions. The only cases that fall outside the range of applicability are the very-small pitch cases for Plate 5, because these cases are insufficiently moderated and also thus have low reactivity. Therefore, the EALF is acceptably within the range of applicability of the benchmarks.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 6.8.2.2 U-235 Number Density The U-235 number density is used as the second trending parameter for the benchmark cases.
The data for all 35 experiments of Set 1 are plotted in Figure 6.8-2. Over the range of applicability, the minimum USL is 0.9240 for Set 1, and 0.9209 for Set 2.
Range of Applicability, Fuel element models: For the optimized fuel element model, the U-235 number densities for plates 1 through 4 and 16 through 19 fall within the range of applicability, while the number densities for plates 5 through 15 exceed the range of applicability (maximum value = 4.22E-03 atom/b-cm). The maximum range of applicability is 3.92E-03 atom/b-cm, so range is exceeded only slightly. If the minimum USL is extrapolated to this larger number density, the minimum USL of 0.9209 does not change. Also, the average U-235 number density for the fuel element is 3.73E-03 atom/b-cm, which is within the allowable range.
Therefore, application of this USL to the fuel element criticality models is considered acceptable.
Range of Applicability, Loose plate models: Of the four plate types modeled, the U-235 number densities for plate type 3 fall within the range of applicability, while the number densities for plate types 5, 8, and 15 exceed the range of applicability (maximum value = 4.29E-03 atom/b-cm). The maximum range of applicability is 3.92E-03 atom/b-cm, so the range is exceeded only slightly. If the minimum USL is extrapolated to this larger number density, the minimum USL of 0.9209 does not change. Therefore, application of this USL to the loose plate basket criticality models is considered acceptable.
6.8.2.3 Channel Width The channel width is used as the third trending parameter for the benchmark cases. The data for all 35 experiments of Set 1 are plotted in Figure 6.8-3. Over the range of applicability, the minimum USL is 0.9225 for Set 1 and 0.9209 for Set 2.
Range of Applicability, Fuel element models: The channel width is fixed at 0.089-in for the most reactive fuel element models, which exceeds the maximum channel width of 0.078-in of the benchmark experiments. However, this parameter is only slightly larger than the maximum benchmark experiment channel width, and was maximized in order to maximize model reactivity. Extrapolation of the USL to the channel width of 0.089-in yields a USL of 0.9208, which is essentially the same as the minimum USL of 0.9209 over the range of applicability.
Therefore, application of this USL to the fuel element criticality models is considered acceptable.
Range of Applicability, Loose plate models: The maximum channel width of the benchmark models is 0.078-in, while the channel width of the most reactive loose plate model is 0.439-in.
Clearly, the loose plate models are well outside the bounds of the benchmark models and extrapolation of the USL would not be appropriate over such a wide range. However, the channel width is directly related to system moderation, and the acceptability of the EALF indicator demonstrates that MCNP is performing acceptably for thermal conditions.
6.8.2.4 H/U-235 Atom Ratio The H/U-235 atom ratio is used as the fourth trending parameter for the benchmark cases. The H/U-235 atom ratio is defined here as the ratio of hydrogen atoms to U-235 atoms in a unit cell.
This parameter is computed by the following equation:
NH*C/(NU235*M) 6-80
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                Rev. 16, May 2021
: where, NH is the hydrogen number density C is the channel width NU235 is the U-235 number density M is the fuel meat width The data for all 35 experiments of Set 1 are plotted in Figure 6.8-4. Over the range of applicability, the minimum USL is 0.9257 for Set 1 and 0.9209 for Set 2.
Range of Applicability, Fuel element models: Using the maximum fuel element plate U-235 number density for the optimized fuel element model, the H/U-235 value may be computed as:
6.687E-02*0.089/(4.224E-03*0.02) = 70.4 Therefore, H/U-235 of the models is acceptably within the range of applicability of the benchmarks.
Range of Applicability, Loose plate models: The H/U-235 atom ratio for the most reactive model may be computed as:
6.687E-02*0.439/(4.2887E-03*0.02) = 342 The maximum H/U-235 atom ratio of the benchmark models is 116.5. Clearly, the loose plate models are well outside the bounds of the benchmark models and extrapolation of the USL would not be appropriate over such a wide range. However, the H/U-235 atom ratio is directly related to system moderation, and the acceptability of the EALF indicator demonstrates that MCNP is performing acceptably for thermal conditions.
6.8.2.5 Pitch The fuel plate pitch is used as the fifth trending parameter for the benchmark cases. The data for all 35 experiments of Set 1 is plotted in Figure 6.8-5. Over the range of applicability, the minimum USL is 0.9225 for Set 1 and 0.9209 for Set 2.
Range of Applicability, Fuel element models: The fuel plate pitch is fixed at 0.128-in for all fuel element models (excluding the pitch for plates 1 and 19, which is slightly bigger because these plates are thicker). This pitch falls within the range of the benchmark experiments.
Range of Applicability, Loose plate models: The maximum pitch of the benchmark models is 0.128-in, while the pitch of the most reactive loose plate model is 0.487-in (1.236 cm). Clearly, the loose plate models are well outside the bounds of the benchmark models and extrapolation of the USL would not be appropriate over such a wide range. However, the pitch is directly related to system moderation, and the acceptability of the EALF indicator demonstrates that MCNP is performing acceptably for thermal conditions.
6.8.2.6 Recommended USL For Set 1, the minimum USL is 0.9225, while for Set 2, the minimum USL is 0.9209. Therefore, the USL is trending lower for Set 2. Note, however, that the average keff = 0.992 for both Set 1 and Set 2. The USL could likely be improved by development of additional benchmark models, 6-81
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 but given the large margins to the most reactive case, the lower value (0.9209) is conservatively selected as the USL for the plate fuel analysis.
6.8.3 Bias Determination for Air Transport Analysis Only the EALF parameter is considered for the air transport benchmark analysis because the other trending parameters used in the plate fuel analysis are generally not applicable to either fast/intermediate systems or the air transport models.
Corresponding data for the fast/intermediate spectrum experiments are summarized in Table 6.8-4. In the plate-fuel benchmarks, the experiments are divided into two sets, a total set of 35 benchmarks (Set 1) and a subset of 17 directly applicable benchmarks (Set 2). These groupings are continued when the fast/intermediate experiments are added:
: 1. Set 3: All 56 experiments (35 plate-fuel benchmark experiments and 21 fast/intermediate benchmark experiments)
: 2. Set 4: 17 directly applicable plate-fuel benchmarks experiments and 21 fast/intermediate benchmark experiments.
: 3. Set 5: 21 fast/intermediate benchmark experiments.
The USL is computed for each of the three sets of experiments, and the results are summarized in Table 6.8-2. When the fast/intermediate spectrum experiments are included, the minimum USL over the range of applicability is 0.9260, 0.9254, and 0.9248 for the three sets considered. These USLs are larger than the minimum USL when the fast/intermediate spectrum experiments are not included (0.9209). Therefore, it is conservative to use a USL of 0.9209 for the air transport analysis.
The range of applicability for EALF over all benchmark experiments is 5.2221x10-8 MeV EALF  0.83695 MeV. The EALF values extracted from the air transport output files are in the range 2.43x10-7 MeV  EALF  8.08x10-3 MeV. Therefore, all air transport cases are within the range of applicability of the benchmark experiments for the EALF parameter.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                    Rev. 16, May 2021 Table 6.8 Benchmark Experiments Utilized Series                                              Title Thermal Plate Fuel Benchmarks HEU-COMP-THERM-022            SPERT III Stainless-Steel-Clad Plate-Type Fuel in Water HEU-MET-THERM-006            SPERT-D Aluminum-Clad Plate-Type Fuel in Water, Dilute Uranyl Nitrate, or Borated Uranyl Nitrate HEU-MET-THERM-022            Advanced Test Reactor: Serpentine Arrangement of Highly Enriched Water-Moderated Uranium-Aluminide Fuel Plates Reflected by Beryllium Intermediate and Fast Benchmarks HEU-MET-FAST-023 (case        Tinkertoy: Unmoderated Uranium Metal (93.2) Arrays With Cylinders
: 16)                          of 10.5 Kg Mass 235 HEU-MET-FAST-032 (case 1)        U(94%) Spheres Surrounded by Natural-Uranium Reflectors HEU-MET-FAST-043 (case 1)    HEU Cylinders Axially Reflected by Steel HEU-MET-FAST-072 (first      Zeus: Fast-Spectrum Critical Assemblies With An Iron - HEU Core configuration)                Surrounded By A Copper Reflector HEU-MET-FAST-076 (cases      Uranium (93.14 235U) Metal Annuli and Cylinders With Thick 5, 8, 12, 16)                Polyethylene Reflectors and/or Internal Polyethylene Moderator HEU-MET-FAST-078 (case 1,    HEU Metal Cylinders, Partially Reflected by Water, Polyethylene,
: 41)                          Lucite, and Paraffin HEU Metal Cylinders With Magnesium, Titanium, Aluminum, Graphite, HEU-MET-FAST-084 (case 7,    Mild Steel, Nickel, Copper, Cobalt, Molybdenum, Natural Uranium,
: 19)                          Tungsten, Beryllium, Aluminum Oxide, Molybdenum Carbide, and Polyethylene Reflectors HEU-MET-FAST-087 (case 1)    Heterogeneous Iron-Diluted HEU Cylinder HEU-MET-INTER-001 (case      The Uranium/Iron Benchmark Assembly: A 235U(93%)/Iron Cylinder
: 1)                            Reflected by Stainless Steel HEU-MET-MIXED-010 (case      Lattices of Oralloy Cubes In Water 3)
HEU-MET-MIXED-011 (case      Oil-Reflected Spheres and Hemispheres of Highly Enriched Uranium 4, 5)                        (93.1% 235U) Metal with Oil or Steel Moderator HEU-MET-MIXED-015 (case      Heterogeneous Cylinder of Highly Enriched Uranium, Polyethylene, and
: 1)                            Titanium With Polyethylene Reflector HEU-MET-THERM-025 (case      Beryllium Moderated Critical Assemblies of Highly Enriched Uranium CA-1)                        (ORNL CA-1 and CA-18)
IEU-MET-FAST-005 (case 1)    Steel-Reflected Spherical Assembly of 235U(36%)
The FR0 Series 4: 20% Enriched "Cylindrical" Uranium Metal Reflected IEU-MET-FAST-021 (case 1) by Natural Uranium 6-83
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                        Rev. 16, May 2021 Table 6.8 USL Results Minimum USL Trending Parameter          Over Range of            Range of (X)              Applicability        Applicability Set 1: 35 Thermal Plate-Fuel Experiments 5.22210E-08 <= X <=
EALF (MeV)                      0.9254              1.58510E-07 U-235 Number Density                            1.84900E-03 <= X <=
0.9240              3.92600E-03 (atom/b-cm) 6.45700E-02 <= X <=
Channel width (in)              0.9225              7.80000E-02 H/U-235                        0.9257          65.100 <= X <= 116.50 Pitch (in)                      0.9225        0.12457 <= X <= 0.12800 Set 2: 17 Directly Applicable Thermal Plate-Fuel Experiments 5.22210E-08 <= X <=
EALF (MeV)                      0.9212              1.58510E-07 U-235 Number Density                            1.84900E-03 <= X <=
0.9209              3.92600E-03 (atom/b-cm) 6.45700E-02 <= X <=
Channel width (in)              0.9209              7.80000E-02 H/U-235                        0.9209          66.0 <= X <= 116.50 Pitch (in)                      0.9209        0.12457 <= X <= 0.12800 Set 3: 35 Thermal Plate-Fuel Experiments plus 21 Fast/Intermediate Experiments EALF (MeV)                      0.9260      5.2221E-08 <= X <= 0.83695 Set 4: 17 Directly Applicable Thermal Plate-Fuel Experiments plus 21 Fast/Intermediate Experiments EALF (MeV)                      0.9254      5.2221E-08 <= X <= 0.83695 Set 5: 21 Fast/Intermediate Experiments EALF (MeV)                      0.9248      2.8058E-06 <= X <= 0.83695 6-84
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                              Rev. 16, May 2021 Table 6.8 Benchmark Experiment Data, Plate Fuels EALF      U-235    Chanel No      Case          k      mcnp bench  total      (MeV)  (atom/b-cm) Width (in) H/U-235      Pitch (in) 1    hct022_c01    0.98895  0.00060 0.0081 0.0081    9.528E-08  3.3155E-03  0.06457    65.1        0.12457 2    hct022_c02    0.98980  0.00061 0.0081 0.0081    9.665E-08  3.3155E-03  0.06457    65.1        0.12457 3    hct022_c03    0.98985  0.00063 0.0081 0.0081    9.809E-08  3.3155E-03  0.06457    65.1        0.12457 4    hct022_c04    0.98856  0.00060 0.0081 0.0081    9.917E-08  3.3155E-03  0.06457    65.1        0.12457 5    hct022_c05    0.98909  0.00063 0.0081 0.0081    9.587E-08  3.3155E-03  0.06457    65.1        0.12457 6    hct022_c06    0.98902  0.00059 0.0081 0.0081    9.840E-08  3.3155E-03  0.06457    65.1        0.12457 7    hct022_c07    0.98963  0.00056 0.0081 0.0081    9.890E-08  3.3155E-03  0.06457    65.1        0.12457 8    hct022_c08    0.98908  0.00057 0.0081 0.0081    9.951E-08  3.3155E-03  0.06457    65.1        0.12457 9    hct022_c09    0.98840  0.00056 0.0081 0.0081    9.589E-08  3.3155E-03  0.06457    65.1        0.12457 10    hct022_c10    0.98845  0.00060 0.0081 0.0081    9.963E-08  3.3155E-03  0.06457    65.1        0.12457 11    hct022_c11    0.98930  0.00060 0.0081 0.0081    1.001E-07  3.3155E-03  0.06457    65.1        0.12457 12    hmt006_c01    0.99240  0.00082 0.0044 0.0045    8.481E-08  1.8490E-03  0.06457    116.5        0.12457 13    hmt006_c02    0.99331  0.00088 0.0040 0.0041    7.044E-08  1.8490E-03  0.06457    116.5        0.12457 14    hmt006_c03    0.99740  0.00072 0.0040 0.0041    6.338E-08  1.8490E-03  0.06457    116.5        0.12457 15    hmt006_c04    0.99282  0.00081 0.0040 0.0041    6.185E-08  1.8490E-03  0.06457    116.5        0.12457 16    hmt006_c05    0.99230  0.00079 0.0040 0.0041    5.852E-08  1.8490E-03  0.06457    116.5        0.12457 17    hmt006_c06    0.99010  0.00071 0.0040 0.0041    5.615E-08  1.8490E-03  0.06457    116.5        0.12457 18    hmt006_c07    0.98783  0.00073 0.0040 0.0041    5.432E-08  1.8490E-03  0.06457    116.5        0.12457 19    hmt006_c08    0.98428  0.00076 0.0040 0.0041    5.245E-08  1.8490E-03  0.06457    116.5        0.12457 20    hmt006_c09    0.98657  0.00072 0.0040 0.0041    5.222E-08  1.8490E-03  0.06457    116.5        0.12457 21    hmt006_c10    0.99885  0.00085 0.0040 0.0041    8.220E-08  1.8490E-03  0.06457    116.5        0.12457 22    hmt006_c11    0.98965  0.00081 0.0040 0.0041    6.236E-08  1.8490E-03  0.06457    116.5        0.12457 23    hmt006_c12    0.99403  0.00070 0.0040 0.0041    5.415E-08  1.8490E-03  0.06457    116.5        0.12457 24    hmt006_c13    1.01283  0.00086 0.0061 0.0062    8.231E-08  1.8490E-03  0.06457    116.5        0.12457 25    hmt006_c14    0.98495  0.00071 0.0040 0.0041    5.715E-08  1.8490E-03  0.06457    116.5        0.12457 (continued) 6-85
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                              Rev. 16, May 2021 Table 6.8 Benchmark Experiment Data, Plate Fuels (concluded)
EALF      U-235    Chanel No      Case          k      mcnp bench  total    (MeV)  (atom/b-cm) Width (in) H/U-235      Pitch (in) 26    hmt006_c15    0.98128  0.00077 0.0040 0.0041    5.654E-08  1.8490E-03  0.06457    116.5        0.12457 27    hmt006_c16    0.99241  0.00078 0.0040 0.0041    6.330E-08  1.8490E-03  0.06457    116.5        0.12457 28    hmt006_c17    0.98934  0.00082 0.0040 0.0041    7.405E-08  1.8490E-03  0.06457    116.5        0.12457 29    hmt006_c18    0.99282  0.00087 0.0040 0.0041    8.003E-08  1.8490E-03  0.06457    116.5        0.12457 30    hmt006_c19    0.99360  0.00068 0.0040 0.0041    5.243E-08  1.8490E-03  0.06457    113.9        0.12457 31    hmt006_c20    0.99275  0.00076 0.0040 0.0041    6.471E-08  1.8490E-03  0.06457    113.7        0.12457 32    hmt006_c21    0.99469  0.00077 0.0040 0.0041    6.917E-08  1.8490E-03  0.06457    113.7        0.12457 33    hmt006_c22    0.99670  0.00080 0.0040 0.0041    7.407E-08  1.8490E-03  0.06457    113.6        0.12457 34    hmt006_c23    1.00132  0.00080 0.0040 0.0041    7.670E-08  1.8490E-03  0.06457    113.5        0.12457 35    hmt022_c01    0.99179  0.00013 0.0035 0.0035    1.585E-07  3.9260E-03  0.078      66.0        0.12800 6-86
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                      Rev. 16, May 2021 Table 6.8 Benchmark Experiment Data, Fast/Intermediate EALF No              Case                    k    mcnp  bench  total    (MeV) 1  HEU-MET-FAST-023, case 16        0.99687 0.00056  0.0052 0.0052    8.113E-01 2  HEU-MET-FAST-032 case 1          1.00427 0.00064  0.0016 0.0017    7.736E-01 3  HEU-MET-FAST-043, case 1          1.00096 0.00059 0.0018  0.0019    7.940E-01 4  HEU-MET-FAST-072, configuration 1 1.00516 0.00065  0.0024 0.0025    4.280E-02 5  HEU-MET-FAST-076, case 5          0.99978 0.00084 0.0006  0.0010    2.560E-03 6  HEU-MET-FAST-076, case 8          0.99814 0.00078 0.0004  0.0009    1.010E-02 7  HEU-MET-FAST-076, case 12        0.99627 0.00076  0.0002 0.0008    5.805E-03 8  HEU-MET-FAST-076, case 16        0.99948 0.00077  0.0005 0.0009    5.410E-03 9  HEU-MET-FAST-078, case 1          0.99209 0.00082 0.0018  0.0020    8.613E-02 10  HEU-MET-FAST-078, case 41        0.99627 0.00061  0.0025 0.0026    8.370E-01 11  HEU-MET-FAST-084, case 19        0.99849 0.00060  0.0019 0.0020    8.016E-01 12  HEU-MET-FAST-084, case 7          1.01966 0.00060 0.0020  0.0021    7.736E-01 13  HEU-MET-FAST-087, case 1          0.99887 0.00060 0.0013  0.0014    7.515E-01 14  HEU-MET-INTER-001, case 1        0.96645 0.00055  0.0026 0.0027    3.908E-02 15  HEU-MET-MIXED-010, case 3        0.97677 0.00072  0.0060 0.0060    1.964E-04 16  HEU-MET-MIXED-011, case 4        1.00179 0.00076  0.0030 0.0031    6.898E-04 17  HEU-MET-MIXED-011, case 5        0.99317 0.00074  0.0041 0.0042    3.530E-04 18  HEU-MET-MIXED-015, case 1        0.99920 0.00080  0.0008 0.0011    9.683E-05 19  HEU-MET-THERM-025, case CA-1      1.01340 0.00109  0.0047 0.0048    2.806E-06 20  IEU-MET-FAST-005, case 1          1.00822 0.00062  0.0023 0.0024    5.738E-01 21  IEU-MET-FAST-021, case 1          1.00471 0.00060 0.00145 0.0016    4.800E-01 6-87
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                      Rev. 16, May 2021 1.015 y = 6344x + 0.9911 1.01                                        2 R = 0.0007 1.005 1
k-eff  0.995 0.99 0.985 0.98 0.975 5.00E-08 7.00E-08 9.00E-08 1.10E-07 1.30E-07 1.50E-07 1.70E-07 EALF (MeV)
Figure 6.8 Benchmark Data Trend for EALF, Set 1 1.015 y = -2.1044x + 0.9965 1.01                                          2 R = 0.0809 1.005 1
k-eff  0.995 0.99 0.985 0.98 0.975 0.0015    0.002    0.0025      0.003        0.0035    0.004 U-235 Number Density (atom/b-cm)
Figure 6.8 Benchmark Data Trend for U-235 Number Density, Set 1 6-88
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                                Rev. 16, May 2021 1.015 1.010 1.005 1.000 y = 0.0186x + 0.9903 2
R = 6E-05 k-eff  0.995 0.990 0.985 0.980 0.975 0.060  0.062    0.064  0.066    0.068    0.070    0.072    0.074  0.076    0.078  0.080 Channel Width (in)
Figure 6.8 Benchmark Data Trend for Channel Width, Set 1 1.015 1.010 y = 6E-05x + 0.9852 1.005 2
R = 0.0832 1.000 k-eff  0.995 0.990 0.985 0.980 0.975 60.0          70.0        80.0            90.0            100.0      110.0          120.0 H/U-235 Figure 6.8 Benchmark Data Trend for H/U-235, Set 1 6-89
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                Rev. 16, May 2021 1.015 1.010 1.005 1.000 y = 0.0733x + 0.9824 2
R = 6E-05 k-eff  0.995 0.990 0.985 0.980 0.975 0.124  0.125  0.125  0.126  0.126  0.127        0.127  0.128  0.128  0.129 Pitch (in)
Figure 6.8 Benchmark Data Trend for Pitch, Set 1 6-90
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                              Rev. 16, May 2021 6.9 Appendix A: Sample Input Files Sample input files are provided for the most reactive NCT array case for both the fuel element payload (Case E23) and the loose plate basket payload (Case LG5).
Case E23 (NA_P030_C89)
ATR 999      0        -320:321:-322:323:-324:325                          imp:n=0 900      0        310 -311 312 -313 24 -25            fill=3          imp:n=1 901      2 -1.0 (311:-310:313:-312:-24:25) 320 -321 322 -323 324 -325                  imp:n=1 c
c        Universe 1: ATR Fuel Element (infinitely long) c 2        3 -2.7              -6 8 9 -10                    u=1  imp:n=1    $ left Al piece 4        3 -2.7              -5 7 9 -10                    u=1  imp:n=1    $ right Al piece 6        10 5.5010E-02      52 14 -13                u=1  imp:n=1    $ plate 1 8        3 -2.7              51 -54    8          #6  u=1  imp:n=1 10      2 -1.00            54 -55    8              u=1  imp:n=1 12      11 5.4998E-02      56 16 -15                u=1  imp:n=1    $ plate 2 14      3 -2.7              55 -58    8          #12  u=1  imp:n=1 16      2 -1.00            58 -59    8              u=1  imp:n=1 18      12 5.4574E-02      60 16 -15                u=1  imp:n=1    $ plate 3 20      3 -2.7              59 -62    8          #18  u=1  imp:n=1 22      2 -1.00            62 -63    8              u=1  imp:n=1 24      13 5.4583E-02      64 16 -15                u=1  imp:n=1    $ plate 4 26      3 -2.7              63 -66    8          #24  u=1  imp:n=1 28      2 -1.00            66 -67    8              u=1  imp:n=1 30      14 5.4115E-02      68 16 -15                u=1  imp:n=1    $ plate 5 32      3 -2.7              67 -70    8          #30  u=1  imp:n=1 34      2 -1.00            70 -71    8              u=1  imp:n=1 36      15 5.4106E-02      72 16 -15                u=1  imp:n=1    $ plate 6 38      3 -2.7              71 -74    8          #36  u=1  imp:n=1 40      2 -1.00            74 -75    8              u=1  imp:n=1 42      16 5.4102E-02      76 16 -15                u=1  imp:n=1    $ plate 7 44      3 -2.7              75 -78    8          #42  u=1  imp:n=1 46      2 -1.00            78 -79    8              u=1  imp:n=1 48      17 5.4098E-02      80 16 -15                u=1  imp:n=1    $ plate 8 50      3 -2.7              79 -82    8          #48  u=1  imp:n=1 52      2 -1.00            82 -83    8              u=1  imp:n=1 54      18 5.4095E-02      84 16 -15                u=1  imp:n=1    $ plate 9 56      3 -2.7              83 -86    8          #54  u=1  imp:n=1 58      2 -1.00            86 -87    8              u=1  imp:n=1 60      19 5.4092E-02      88 16 -15                u=1  imp:n=1    $ plate 10 62      3 -2.7              87 -90    8          #60  u=1  imp:n=1 64      2 -1.00            90 -91    8              u=1  imp:n=1 66      20 5.4089E-02      92 16 -15                u=1  imp:n=1    $ plate 11 68      3 -2.7              91 -94    8          #66  u=1  imp:n=1 70      2 -1.00            94 -95    8              u=1  imp:n=1 72      21 5.4086E-02      96 16 -15                u=1  imp:n=1    $ plate 12 74      3 -2.7              95 -98    8          #72  u=1  imp:n=1 76      2 -1.00            98 -99    8              u=1  imp:n=1 78      22 5.4083E-02      100 -101 15              u=1  imp:n=1    $ plate 13 80      3 -2.7              99 -102    8        #78  u=1  imp:n=1 82      2 -1.00            102 -103    8            u=1  imp:n=1 84      23 5.4081E-02      104 -105 15              u=1  imp:n=1    $ plate 14 6-91
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                Rev. 16, May 2021 86      3 -2.7          103 -106  8    #84 u=1 imp:n=1 88      2 -1.00        106 -107  8          u=1 imp:n=1 90      24 5.4075E-02 108 -109 15            u=1 imp:n=1 $ plate 15 92      3 -2.7          107 -110  8    #90 u=1 imp:n=1 94      2 -1.00        110 -111  8          u=1 imp:n=1 96      25 5.4544E-02 112 -113 15            u=1 imp:n=1 $ plate 16 98      3 -2.7          111 -114  8    #96 u=1 imp:n=1 100    2 -1.00        114 -115  8          u=1 imp:n=1 102    26 5.4544E-02 116 -117 15            u=1 imp:n=1 $ plate 17 104    3 -2.7          115 -118  8    #102 u=1 imp:n=1 106    2 -1.00        118 -119  8          u=1 imp:n=1 108    27 5.4949E-02 120 -121 17            u=1 imp:n=1 $ plate 18 110    3 -2.7          119 -122  8    #108 u=1 imp:n=1 112    2 -1.00        122 -123  8          u=1 imp:n=1 114    28 5.4967E-02 124 -125 13            u=1 imp:n=1 $ plate 19 116    3 -2.7          123 -126  8    #114 u=1 imp:n=1 c 122    2 -1.00        6:5:-9:10:9 8 -7:126 8 -7 u=1 imp:n=1 120    2 -1.00        126 8 -7            u=1 imp:n=1 $ above 19 121    2 -1.00        9 8 -7              u=1 imp:n=1 $ below 1 122    5 -0.737        5 -11 9 -10              u=1 imp:n=1 $ right neoprene 123    5 -0.737    -12 6 9 -10                  u=1 imp:n=1 $ left neoprene 125    2 -1.0        12:11:-9:10                u=1 imp:n=1 c
c      Universe 20: ATR with pipe (center) c 200    0  26 22 -23:26 -20 22 -28:27 -21  22 -28  trcl=1 fill=1 u=20 imp:n=1 201    2 -0.3    #200 -200                    u=20 imp:n=1 $ between ATR/pipe 202    4 -7.94    200 -201                    u=20 imp:n=1 $ pipe 203    6 -0.096    201 -203 250 -251 252 -253  u=20 imp:n=1 $ insulation 204    0          203 250 -251 252 -253      u=20 imp:n=1 $ insulation to tube 205    4 -7.94    -250:251:-252:253          u=20 imp:n=1 $ tube to inf c
c      Universe 21: ATR with pipe (down) c 210    0 26 22 -23:26 -20 22 -28:27 -21  22 -28  trcl=2 fill=1 u=21 imp:n=1 211    2 -0.3    #210 -200                    u=21 imp:n=1 $ between ATR/pipe 212    4 -7.94    200 -201                    u=21 imp:n=1 $ pipe 213    6 -0.096    201 -203 250 -251 252 -253  u=21 imp:n=1 $ insulation 214    0          203 250 -251 252 -253      u=21 imp:n=1 $ insulation to tube 215    4 -7.94    -250:251:-252:253          u=21 imp:n=1 $ tube to inf c
c      Universe 22: ATR with pipe (up) c 220    0 26 22 -23:26 -20 22 -28:27 -21  22 -28  trcl=3 fill=1 u=22 imp:n=1 221    2 -0.3    #220 -200                    u=22 imp:n=1 $ between ATR/pipe 222    4 -7.94    200 -201                    u=22 imp:n=1 $ pipe 223    6 -0.096    201 -203 250 -251 252 -253  u=22 imp:n=1 $ insulation 224    0          203 250 -251 252 -253      u=22 imp:n=1 $ insulation to tube 6-92
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                Rev. 16, May 2021 225    4 -7.94    -250:251:-252:253          u=22 imp:n=1 $ tube to inf c
c      Universe 23: ATR with pipe (right) c 230    0  26 22 -23:26 -20 22 -28:27 -21 22 -28 trcl=4 fill=1 u=23 imp:n=1 231    2 -0.3    #230 -200                  u=23 imp:n=1 $  between ATR/pipe 232    4 -7.94    200 -201                  u=23 imp:n=1 $  pipe 233    6 -0.096    201 -203 250 -251 252 -253 u=23 imp:n=1 $  insulation 234    0          203 250 -251 252 -253      u=23 imp:n=1 $  insulation to tube 235    4 -7.94    -250:251:-252:253          u=23 imp:n=1 $  tube to inf c
c      Universe 24: ATR with pipe (left) c 240    0 26 22 -23:26 -20 22 -28:27 -21 22 -28 trcl=5 fill=1 u=24 imp:n=1 241    2 -0.3    #240 -200                  u=24 imp:n=1 $  between ATR/pipe 242    4 -7.94    200 -201                  u=24 imp:n=1 $  pipe 243    6 -0.096    201 -203 250 -251 252 -253 u=24 imp:n=1 $  insulation 244    0          203 250 -251 252 -253      u=24 imp:n=1 $  insulation to tube 245    4 -7.94    -250:251:-252:253          u=24 imp:n=1 $  tube to inf c
c      Universe 25: ATR with pipe (up right) c 250    0 26 22 -23:26 -20 22 -28:27 -21 22 -28 trcl=6 fill=1 u=25 imp:n=1 251    2 -0.3    #250 -200                  u=25 imp:n=1 $  between ATR/pipe 252    4 -7.94    200 -201                  u=25 imp:n=1 $  pipe 253    6 -0.096    201 -203 250 -251 252 -253 u=25 imp:n=1 $  insulation 254    0          203 250 -251 252 -253      u=25 imp:n=1 $  insulation to tube 255    4 -7.94    -250:251:-252:253          u=25 imp:n=1 $  tube to inf c
c      Universe 26: ATR with pipe (up left) c 260    0 26 22 -23:26 -20 22 -28:27 -21 22 -28 trcl=7 fill=1 u=26 imp:n=1 261    2 -0.3    #260 -200                  u=26 imp:n=1 $  between ATR/pipe 262    4 -7.94    200 -201                  u=26 imp:n=1 $  pipe 263    6 -0.096    201 -203 250 -251 252 -253 u=26 imp:n=1 $  insulation 264    0          203 250 -251 252 -253      u=26 imp:n=1 $  insulation to tube 265    4 -7.94    -250:251:-252:253          u=26 imp:n=1 $  tube to inf c
c      Universe 27: ATR with pipe (down right) c 270    0 26 22 -23:26 -20 22 -28:27 -21 22 -28 trcl=8 fill=1 u=27 imp:n=1 271    2 -0.3    #270 -200                  u=27 imp:n=1 $  between ATR/pipe 272    4 -7.94    200 -201                  u=27 imp:n=1 $  pipe 6-93
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                Rev. 16, May 2021 273    6 -0.096    201 -203 250 -251 252 -253 u=27 imp:n=1 $ insulation 274    0          203 250 -251 252 -253      u=27 imp:n=1 $ insulation to tube 275    4 -7.94    -250:251:-252:253          u=27 imp:n=1 $ tube to inf c
c      Universe 28: ATR with pipe (down left) c 280    0  26 22 -23:26 -20 22 -28:27 -21  22 -28  trcl=9 fill=1 u=28 imp:n=1 281    2 -0.3    #280 -200                    u=28 imp:n=1 $ between ATR/pipe 282    4 -7.94    200 -201                    u=28 imp:n=1 $ pipe 283    6 -0.096    201 -203 250 -251 252 -253  u=28 imp:n=1 $ insulation 284    0          203 250 -251 252 -253      u=28 imp:n=1 $ insulation to tube 285    4 -7.94    -250:251:-252:253          u=28 imp:n=1 $ tube to inf c
c      Universe 3: Array of Packages c
300  0    -300 301 -302 303 imp:n=1 u=3 lat=1  fill=-4:4 -4:4 0:0 25 25 25 25 22 26 26 26 26 25 25 25 25 22 26 26 26 26 25 25 25 25 22 26 26 26 26 25 25 25 25 22 26 26 26 26 23 23 23 23 20 24 24 24 24 27 27 27 27 21 28 28 28 28 27 27 27 27 21 28 28 28 28 27 27 27 27 21 28 28 28 28 27 27 27 27 21 28 28 28 28 5      p 2.4142136  -1  0 -0.2665911 $ right Al outer 6      p -2.4142136 -1  0 -0.2665911 $ left Al outer 7      p 2.4142136  -1  0 -1.474587  $ right Al inner 8      p -2.4142136 -1  0 -1.474587  $ left Al inner 9      cz 7.52856                    $ Al boundary 10      cz 14.015466                  $ Al boundary 11      p 2.4142136  -1 0  0.563076  $ right neoprene 12      p -2.4142136 -1 0  0.563076  $ left neoprene c
13      p 2.4142136 -1 0 -2.4370013 $  plate 1 & 19 meat 14      p -2.4142136 -1 0 -2.4370013 $  plate 1 & 19 meat 15      p 2.4142136 -1 0 -1.7732672 $  plate 2-17 meat 16      p -2.4142136 -1 0 -1.7732672 $  plate 2-17 meat 17      p 2.4142136 -1 0 -1.9060140 $  plate 18 meat 18      p -2.4142136 -1 0 -1.9060140 $  plate 18 meat c
20      p 2.4142136  -1 0    0.6      $ right u0 boundary 21      p -2.4142136 -1 0    0.6      $ left u0 boundary 22      cz 7.51                      $ u0 boundary 23      cz 14.02                      $ u0 boundary 24      pz -60.96                    $ bottom of fuel 25      pz 60.96                      $ top of fuel (48")
26      p 2.4142136  -1 0  0.0      $ neoprene notch 27      p -2.4142136 -1 0  0.0      $ neoprene notch 28      cz 13.9                      $ neoprene notch c
51      cz 7.67207  $ fuel plate 1 (0.089) 6-94
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report            Rev. 16, May 2021 52    cz 7.7343 53    cz 7.7851 54    cz 7.84733 c
55    cz 8.07339  $ fuel plate 2 56    cz 8.09752 57    cz 8.14832 58    cz 8.17245 c
59    cz 8.39851  $ fuel plate 3 60    cz 8.42264 61    cz 8.47344 62    cz 8.49757 c
63    cz 8.72363  $ fuel plate 4 64    cz 8.74776 65    cz 8.79856 66    cz 8.82269 c
67    cz 9.04875  $ fuel plate 5 68    cz 9.07288 69    cz 9.12368 70    cz 9.14781 c
71    cz 9.37387  $ fuel plate 6 72    cz 9.398 73    cz 9.4488 74    cz 9.47293 c
75    cz 9.69899  $ fuel plate 7 76    cz 9.72312 77    cz 9.77392 78    cz 9.79805 c
79    cz 10.02411  $ fuel plate 8 80    cz 10.04824 81    cz 10.09904 82    cz 10.12317 c
83    cz 10.34923  $ fuel plate 9 84    cz 10.37336 85    cz 10.42416 86    cz 10.44829 c
87    cz 10.67435  $ fuel plate 10 88    cz 10.69848 89    cz 10.74928 90    cz 10.77341 c
91    cz 10.99947  $ fuel plate 11 92    cz 11.0236 93    cz 11.0744 94    cz 11.09853 c
95    cz 11.32459  $ fuel plate 12 96    cz 11.34872 97    cz 11.39952 6-95
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                  Rev. 16, May 2021 98    cz 11.42365 c
99      cz 11.64971    $ fuel plate 13 100    cz 11.67384 101    cz 11.72464 102    cz 11.74877 c
103    cz 11.97483    $ fuel plate 14 104    cz 11.99896 105    cz 12.04976 106    cz 12.07389 c
107    cz 12.29995    $ fuel plate 15 108    cz 12.32408 109    cz 12.37488 110    cz 12.39901 c
111    cz 12.62507    $ fuel plate 16 112    cz 12.6492 113    cz 12.7 114    cz 12.72413 c
115    cz 12.95019    $ fuel plate 17 116    cz 12.97432 117    cz 13.02512 118    cz 13.04925 c
119    cz 13.27531    $ fuel plate 18 120    cz 13.29944 121    cz 13.35024 122    cz 13.37437 c
123    cz 13.60043    $ fuel plate 19 (0.089) 124    cz 13.68806 125    cz 13.73886 126    cz 13.82649 c
200    cz 7.3838 $ IR pipe 201    cz 7.6581 $ OR pipe 202    cz 38.1    $ 12" water 203    cz 10.1981 $ 1" insulation c
250    px  -9.6032 $ square tube 251    px    9.6032 252    py  -9.6032 253    py    9.6032 c
300    px  10.033 $ lattice surfaces/sq. tube 301    px -10.033 302    py  10.033 303    py -10.033 310    px -90.297 $ 9x9 bounds 311    px  90.297 312    py -90.297 313    py  90.297 320    px -120.777 $ outer bounds 321    px  120.777 6-96
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                      Rev. 16, May 2021 322    py -120.777 323    py 120.777 324    pz -91.44 325    pz 91.44 m2    1001.62c 2            $ water 8016.62c 1 mt2    lwtr.60t m3    13027.62c 1            $ Al m4    6000.66c    -0.08      $ SS-304 14000.60c -1.0 15031.66c -0.045 24000.50c -19.0 25055.62c -2.0 26000.55c -68.375 28000.50c -9.5 m5    1001.62c -0.056920    $ neoprene 6000.66c -0.542646 c      17000.66c -0.400434 m6    13027.62c -26.5        $ insulation material 14000.60c -23.4 8016.62c    -50.2 m10    92234.69c 1.7026E-05  $ fuel plate 1 92235.69c 2.6560E-03 92236.69c 9.8475E-06 92238.69c 1.4089E-04 13027.62c 5.2187E-02 c      total      5.5010E-02 m11    92234.69c 1.7156E-05  $ fuel plate 2 92235.69c 2.6763E-03 92236.69c 9.9226E-06 92238.69c 1.4196E-04 13027.62c 5.2153E-02 c      total      5.4998E-02 m12    92234.69c 2.1711E-05  $ fuel plate 3 92235.69c 3.3869E-03 92236.69c 1.2557E-05 92238.69c 1.7966E-04 13027.62c 5.0974E-02 c      total      5.4574E-02 m13    92234.69c 2.1618E-05  $ fuel plate 4 92235.69c 3.3724E-03 92236.69c 1.2503E-05 92238.69c 1.7889E-04 13027.62c 5.0998E-02 c      total      5.4583E-02 m14    92234.69c 2.6648E-05  $ fuel plate 5 92235.69c 4.1571E-03 92236.69c 1.5413E-05 92238.69c 2.2051E-04 13027.62c 4.9696E-02 c      total      5.4115E-02 m15    92234.69c 2.6746E-05  $ fuel plate 6 92235.69c 4.1724E-03 92236.69c 1.5470E-05 92238.69c 2.2132E-04 13027.62c 4.9670E-02 6-97
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                Rev. 16, May 2021 c      total      5.4106E-02 m16    92234.69c  2.6790E-05  $ fuel plate 7 92235.69c  4.1791E-03 92236.69c  1.5495E-05 92238.69c  2.2168E-04 13027.62c  4.9659E-02 c      total      5.4102E-02 m17    92234.69c  2.6830E-05  $ fuel plate 8 92235.69c  4.1854E-03 92236.69c  1.5518E-05 92238.69c  2.2201E-04 13027.62c  4.9649E-02 c      total      5.4098E-02 m18    92234.69c  2.6867E-05  $ fuel plate 9 92235.69c  4.1911E-03 92236.69c  1.5539E-05 92238.69c  2.2232E-04 13027.62c  4.9639E-02 c      total      5.4095E-02 m19    92234.69c  2.6901E-05  $ fuel plate 10 92235.69c  4.1965E-03 92236.69c  1.5559E-05 92238.69c  2.2260E-04 13027.62c  4.9630E-02 c      total      5.4092E-02 m20    92234.69c  2.6933E-05  $ fuel plate 11 92235.69c  4.2015E-03 92236.69c  1.5577E-05 92238.69c  2.2287E-04 13027.62c  4.9622E-02 c      total      5.4089E-02 m21    92234.69c  2.6963E-05  $ fuel plate 12 92235.69c  4.2061E-03 92236.69c  1.5595E-05 92238.69c  2.2311E-04 13027.62c  4.9614E-02 c      total      5.4086E-02 m22    92234.69c  2.6990E-05  $ fuel plate 13 92235.69c  4.2105E-03 92236.69c  1.5611E-05 92238.69c  2.2334E-04 13027.62c  4.9607E-02 c      total      5.4083E-02 m23    92234.69c  2.7017E-05  $ fuel plate 14 92235.69c  4.2145E-03 92236.69c  1.5626E-05 92238.69c  2.2356E-04 13027.62c  4.9600E-02 c      total      5.4081E-02 m24    92234.69c  2.7077E-05  $ fuel plate 15 92235.69c  4.2239E-03 92236.69c  1.5661E-05 92238.69c  2.2406E-04 13027.62c  4.9585E-02 c      total      5.4075E-02 m25    92234.69c  2.2037E-05  $ fuel plate 16 92235.69c  3.4377E-03 6-98
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                Rev. 16, May 2021 92236.69c 1.2746E-05 92238.69c 1.8235E-04 13027.62c 5.0889E-02 c      total    5.4544E-02 m26    92234.69c 2.2037E-05 $ fuel plate 17 92235.69c 3.4377E-03 92236.69c 1.2745E-05 92238.69c 1.8235E-04 13027.62c 5.0889E-02 c      total    5.4544E-02 m27    92234.69c 1.7683E-05 $ fuel plate 18 92235.69c 2.7586E-03 92236.69c 1.0228E-05 92238.69c 1.4633E-04 13027.62c 5.2016E-02 c      total    5.4949E-02 m28    92234.69c 1.7487E-05 $ fuel plate 19 92235.69c 2.7279E-03 92236.69c 1.0114E-05 92238.69c 1.4470E-04 13027.62c 5.2067E-02 c      total    5.4967E-02 c
*tr1    0 -10.8 0                        $ base to center
*tr2    0 7.9 0    180 90 90 90 180 90    $ down
*tr3    0 -7.9 0                          $ up
*tr4    -7.9 0 0    90 180 90 0 90 90      $ right
*tr5    7.9 0 0    90 0 90 180 90 90      $ left
*tr6    -5.6 -5.6  0 45 135 90 45 45 90    $ up/right
*tr7    5.6 -5.6  0 45 45 90 135 45 90    $ up/left
*tr8    -5.6  5.6 0 135 135 90 45 135 90 $  down/right
*tr9    5.6  5.6 0 135 45 90 135 135 90 $  down/left c
mode  n kcode  2500 1.0 50 250 sdef    x=d1 y=d2 z=d3 si1    -90 90 sp1    0 1 si2    -90 90 sp2    0 1 si3    -60 60 sp3    0 1 Case LG5 (NA_N5P050)
ATR 999    0      -320:321:-322:323:-324:325                imp:n=0 900    0      310 -311 312 -313 24 -25      fill=3      imp:n=1 901    2 -1.0 (311:-310:313:-312:-24:25) 320 -321 322 -323 324 -325 imp:n=1 c
c      Universe 5: Plate 5 c
500    14 5.4037E-02    500 -501 502 -503            u=5 imp:n=1 $ fuel meat 501    3 -2.7          (-500:501:-502:503) 510 -511 512 -513 u=5 imp:n=1 $
cladding 502    2 -1.0          -510:511:-512:513                      u=5 imp:n=1 $
water 6-99
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                Rev. 16, May 2021 c
c      Universe 6: Lattice c
600    0  -531 530 lat=1 fill=-2:2 0:0 0:0 5
5(0 -0.2 0) 5 5(0 0.2 0) 5 imp:n=1 u=6 c
c      Universe 4: Plates and basket (no pipe) c 400    0          520 -521 522 -523    fill=6(0 0 0)    imp:n=1 u=4 $ fuel lattice 401    2 -0.5    (-520:521:-522:523) 400 -401 402 -403 imp:n=1 u=4 $ water between fuel and basket 402    3 -2.7    -400:401:-402:403          imp:n=1 u=4 $ basket (to infinity) c c      Universe 20: Plates with pipe (center) c 200    0        410 -411 412 -413          fill=4 imp:n=1 u=20 $ fuel/basket 201    2 -0.5    #200 -200              imp:n=1 u=20 $ water between basket and tube 202    4 -7.94  200 -201                          imp:n=1 u=20 $ tube 203    6 -0.096    201 -203 250 -251 252 -253      imp:n=1 u=20 $ insulation 204    0          203 250 -251 252 -253          imp:n=1 u=20 $ insulation to tube 205    4 -7.94    -250:251:-252:253              imp:n=1 u=20 $ tube to inf c
c      Universe 21: Plates with pipe (down) c 210    0        410 -411 412 -413 trcl=2 fill=4 imp:n=1 u=21 $ fuel/basket 211    2 -0.5    #210 -200              imp:n=1 u=21 $ water between basket and tube 212    4 -7.94  200 -201                          imp:n=1 u=21 $ tube 213    6 -0.096    201 -203 250 -251 252 -253      imp:n=1 u=21 $ insulation 214    0          203 250 -251 252 -253          imp:n=1 u=21 $ insulation to tube 215    4 -7.94    -250:251:-252:253              imp:n=1 u=21 $ tube to inf c
c      Universe 22: Plates with pipe (up) c 220    0        410 -411 412 -413 trcl=3 fill=4 imp:n=1 u=22 $ fuel/basket 221    2 -0.5    #220 -200              imp:n=1 u=22 $ water between basket and tube 222    4 -7.94  200 -201                          imp:n=1 u=22 $ tube 223    6 -0.096    201 -203 250 -251 252 -253      imp:n=1 u=22 $ insulation 224    0          203 250 -251 252 -253          imp:n=1 u=22 $ insulation to tube 225    4 -7.94    -250:251:-252:253              imp:n=1 u=22 $ tube to inf c
c      Universe 23: Plates with pipe (right) c 230    0        410 -411 412 -413 trcl=4 fill=4 imp:n=1 u=23 $ fuel/basket 231    2 -0.5    #230 -200              imp:n=1 u=23 $ water between basket and tube 6-100
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                Rev. 16, May 2021 232    4 -7.94  200 -201                        imp:n=1 u=23 $ tube 233    6 -0.096    201 -203 250 -251 252 -253    imp:n=1 u=23 $ insulation 234    0          203 250 -251 252 -253          imp:n=1 u=23 $ insulation to tube 235    4 -7.94    -250:251:-252:253              imp:n=1 u=23 $ tube to inf c
c      Universe 24: Plates with pipe (left) c 240    0        410 -411 412 -413 trcl=5 fill=4 imp:n=1 u=24 $ fuel/basket 241    2 -0.5    #240 -200              imp:n=1 u=24 $ water between basket and tube 242    4 -7.94  200 -201                        imp:n=1 u=24 $ tube 243    6 -0.096    201 -203 250 -251 252 -253    imp:n=1 u=24 $ insulation 244    0          203 250 -251 252 -253          imp:n=1 u=24 $ insulation to tube 245    4 -7.94    -250:251:-252:253              imp:n=1 u=24 $ tube to inf c
c      Universe 25: Plates with pipe (up right) c 250    0        410 -411 412 -413 trcl=6 fill=4 imp:n=1 u=25 $ fuel/basket 251    2 -0.5    #250 -200            imp:n=1 u=25 $ water between basket and tube 252    4 -7.94  200 -201                        imp:n=1 u=25 $ tube 253    6 -0.096    201 -203 250 -251 252 -253    imp:n=1 u=25 $ insulation 254    0          203 250 -251 252 -253          imp:n=1 u=25 $ insulation to tube 255    4 -7.94    -250:251:-252:253              imp:n=1 u=25 $ tube to inf c
c      Universe 26: Plates with pipe (up left) c 260    0        410 -411 412 -413 trcl=7 fill=4 imp:n=1 u=26 $ fuel/basket 261    2 -0.5    #260 -200          imp:n=1 u=26 $ water between basket and tube 262    4 -7.94  200 -201                        imp:n=1 u=26 $ tube 263    6 -0.096    201 -203 250 -251 252 -253    imp:n=1 u=26 $ insulation 264    0          203 250 -251 252 -253          imp:n=1 u=26 $ insulation to tube 265    4 -7.94    -250:251:-252:253              imp:n=1 u=26 $ tube to inf c
c      Universe 27: Plates with pipe (down right) c 270    0        410 -411 412 -413 trcl=8 fill=4 imp:n=1 u=27 $ fuel/basket 271    2 -0.5    #270 -200            imp:n=1 u=27 $ water between basket and tube 272    4 -7.94  200 -201                        imp:n=1 u=27 $ tube 273    6 -0.096    201 -203 250 -251 252 -253    imp:n=1 u=27 $ insulation 274    0          203 250 -251 252 -253          imp:n=1 u=27 $ insulation to tube 275    4 -7.94    -250:251:-252:253              imp:n=1 u=27 $ tube to inf c
c      Universe 28: Plates with pipe (down left) c 280    0        410 -411 412 -413 trcl=9 fill=4 imp:n=1 u=28 $ fuel/basket 281    2 -0.5    #280 -200            imp:n=1 u=28 $ water between basket and tube 282    4 -7.94  200 -201                        imp:n=1 u=28 $ tube 283    6 -0.096    201 -203 250 -251 252 -253    imp:n=1 u=28 $ insulation 6-101
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                              Rev. 16, May 2021 284    0          203 250 -251 252 -253        imp:n=1 u=28 $ insulation to tube 285    4 -7.94    -250:251:-252:253            imp:n=1 u=28 $ tube to inf c
c      Universe 3: Array of Packages c
300  0    -300 301 -302 303 imp:n=1 u=3 lat=1 fill=-4:4 -4:4 0:0 25 25 25 25 22 26 26 26 26 25 25 25 25 22 26 26 26 26 25 25 25 25 22 26 26 26 26 25 25 25 25 22 26 26 26 26 23 23 23 23 20 24 24 24 24 27 27 27 27 21 28 28 28 28 27 27 27 27 21 28 28 28 28 27 27 27 27 21 28 28 28 28 27 27 27 27 21 28 28 28 28 24      pz -60.96                    $ bottom of fuel 25      pz 60.96                    $ top of fuel (48")
c 200      cz 7.3838 $ IR pipe 201      cz 7.6581 $ OR pipe 203      cz 10.1981 $ 1" insulation c
250      px  -9.6032 $ square tube 251      px  9.6032 252      py  -9.6032 253      py  9.6032 c
300      px  10.033 $ lattice surfaces/sq. tube 301      px -10.033 302      py  10.033 303      py -10.033 310      px -90.297 $ 9x9 bounds 311      px  90.297 312      py -90.297 313      py  90.297 320      px -120.777 $ outer bounds 321      px  120.777 322      py -120.777 323      py  120.777 324      pz -91.44 325      pz  91.44 c
400      px -5.7912  $ inner basket surfaces 401      px  5.7912 402      py -2.1336 403      py  2.1336 410      px -6.1214  $ outer basket surfaces 411      px  6.1214 412      py -2.4638 413      py  2.4638 c
500      px -5.7873 $ fuel meat 501      px 5.7873 502      py -0.0254 503      py 0.0254 6-102
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                              Rev. 16, May 2021 510    px -5.79    $ fuel cladding 511    px  5.79 512    py -0.06096 513    py  0.06096 520    px -5.791 $ array boundary 521    px  5.791 522    py -2.13296 523    py  2.13296 530    py -0.518      $ lattice bounds 531    py  0.518 m2    1001.62c  2          $ water 8016.62c  1 mt2    lwtr.60t m3    13027.62c  1          $ Al m4    6000.66c    -0.08      $ SS-304 14000.60c  -1.0 15031.66c  -0.045 24000.50c  -19.0 25055.62c  -2.0 26000.55c  -68.375 28000.50c  -9.5 m6    13027.62c  -26.5      $ insulation material 14000.60c  -23.4 8016.62c    -50.2 m14    92234.69c  2.7492E-05  $ plate 5 92235.69c  4.2887E-03 92236.69c  1.5901E-05 92238.69c  2.2749E-04 13027.62c  4.9477E-02 c      total      5.4037E-02 c
*tr2    0 -1.6 0                                $ down
*tr3    0    1.6 0                              $ up
*tr4    1.6 0    0        90 180 90 0 90 90    $ right
*tr5  -1.6 0    0        90 0 90 180 90 90    $ left
*tr6    1.13 1.13  0    45 135 90 45 45 90    $ up/right
*tr7  -1.13 1.13  0    45 45 90 135 45 90    $ up/left
*tr8    1.13 -1.13  0    135 135 90 45 135 90  $ down/right
*tr9  -1.13 -1.13  0    135 45 90 135 135 90  $ down/left c
mode  n kcode 2500 1.0 50 250 sdef  x=d1 y=d2 z=d3 si1    -90 90 sp1    0 1 si2    -90 90 sp2    0 1 si3    -60 60 sp3    0 1 6-103
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                        Rev. 16, May 2021 6.10 Appendix B: Criticality Analysis for MIT and MURR Fuel The ATR FFSC may be utilized to transport MIT fuel and MURR fuel. Both of these fuels are high-enriched plate-type fuels similar to the ATR fuel analyzed in this chapter, although the fuel geometries are different. The following analyses demonstrate that the ATR FFSC with the MIT and MURR fuel complies with the requirements of 10 CFR &sect;71.55 and &sect;71.59. Based on a 5x5 array of damaged packages, the Criticality Safety Index (CSI), per 10 CFR &sect;71.59, is 4.0.
6.10.1 Description of Criticality Design 6.10.1.1 Design Features Important for Criticality No special design features are required to maintain criticality safety. No poisons are utilized in the package. The MURR and MIT fuel handling enclosures (FHEs) restrict postulated fuel element pitch expansion under hypothetical accident conditions. In addition, the separation provided by the packaging (outer flat-to-flat dimension of 7.9-in), along with the limit on the number of packages per shipment, is sufficient to maintain criticality safety.
6.10.1.2 Summary Table of Criticality Evaluation The upper subcritical limit (USL) for ensuring that the ATR FFSC (single package or package array) is acceptably subcritical, is:
USL = 0.9209 The package is considered to be acceptably subcritical if the computed ksafe (ks), which is defined as keffective (keff) plus twice the statistical uncertainty (), is less than or equal to the USL, or:
ks = keff + 2  USL The USL is determined on the basis of a benchmark analysis and incorporates the combined effects of code computational bias, the uncertainty in the bias based on both benchmark-model and computational uncertainties, and an administrative margin. The results of the benchmark analysis indicate that the USL is adequate to ensure subcriticality of the package.
The packaging design is shown to meet the requirements of 10 CFR 71.55(b). Moderation by water in the most reactive credible extent is utilized in both the normal conditions of transport (NCT) and hypothetical accident conditions of transport (HAC) analyses. In the single package NCT models, full-density water fills the accessible cavity, while in the single package HAC models, full-density water fills all cavities. In the NCT fuel element models, the fuel element is modeled as undamaged, although the most reactive credible configuration is utilized by maximizing the gap between the fuel plates. Maximizing this gap maximizes the moderation and hence the reactivity because the system is undermoderated. In the HAC fuel element models, a damaged fuel element is assumed, and the fuel element pitch is allowed to expand until constrained by the FHE, which maximizes moderation. In all single package models, 12-in of water reflection is utilized.
In the NCT and HAC array cases, partial moderation is considered to maximize array interaction effects. A 9x9x1 array is utilized for the NCT array, while a 5x5x1 array is utilized in the HAC array. In all array models, 12-in of water reflection are utilized.
6-104
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                Rev. 16, May 2021 The maximum results of the criticality calculations are summarized in Table 6.10-1. The MURR fuel is significantly more reactive than the MIT fuel. The maximum calculated ks is 0.85881, which occurs for the optimally moderated MURR HAC array case. In this case, the FHE is moderated with full-density water, the inner tube (outside the FHE) is moderated with 0.8 g/cm3 water, and void is modeled between the insulation and outer tube.
6.10.1.3 Criticality Safety Index The criticality safety index of 4.0 for MIT and MURR fuel is unchanged from the value provided in Section 6.1.3, Criticality Safety Index.
Table 6.10 Summary of Criticality Evaluation MURR                MIT Normal Conditions of Transport (NCT)
Case                          ks                ks Single Unit Maximum                  0.44807            0.36978 9x9 Array Maximum                  0.85643            0.65658 Hypothetical Accident Conditions (HAC)
Case                          ks                ks Single Unit Maximum                  0.54584            0.43666 5x5 Array Maximum                  0.85881            0.67309 USL = 0.9209 6.10.2 Fissile Material Contents The package can accommodate either one MURR or one MIT fuel element. The geometry and composition of these fuel elements are described in the following sections.
6.10.2.1 MURR Fuel Element Each MURR element contains up to 785 g U-235, enriched up to 94 wt.%. The weight percents of the remaining uranium isotopes are 1.2 wt.% U-234, 0.7 wt.% U-236, and 5.0-7.0 wt.%
U-238. Each fuel element contains 24 curved fuel plates. Fuel plate 1 has the smallest radius, while fuel plate 24 has the largest radius, as shown in Figure 6.10-1 and Figure 6.10-3. The fuel meat is a mixture of uranium metal and aluminum, while the cladding and structural materials are an aluminum alloy.
The geometry of the fuel element is defined in Figure 6.10-1. Each fuel plate is nominally 0.05-in thick, with a thickness tolerance of +/-0.002-in. The fuel meat is nominally 0.02-in thick, and the cladding is nominally 0.015-in thick. The plate cladding material is aluminum. Fuel element side plates are fabricated of ASTM B 209, aluminum alloy 6061-T6 or 6061-T651. These fuel element side plates have a minimum thickness of 0.145-in. The channel width between the plates is 0.080 +/- 0.008-in. This tolerance represents average and not localized channel width.
6-105
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                    Rev. 16, May 2021 For an actual fuel element, the channel width may exceed this tolerance in localized areas. The maximum local channel width is 0.090-in.
The arc length of the fuel meat changes from plate to plate. Reference fuel meat arc length and inner radius dimensions for each plate are provided in Table 6.10-2. The active fuel length ranges from 23.25-in to 24.75-in as illustrated in Figure 6.10-1.
It is necessary to determine the number densities of the fuel meat, which are the same for all fuel plates. To determine the number densities of the fuel meat, it is first necessary to compute the volume of the fuel meat. The volume of the fuel meat for each plate is the arc length of the meat (nominal + 0.065-in) multiplied by the active fuel length (24.0-in) and meat thickness (0.02-in).
The active fuel length and meat thickness are modeled at nominal values in all final (i.e., non-parametric) fuel element models, and the use of these dimensions is justified in Section 6.10.4.1.2, HAC Single Package Configuration. It is demonstrated that reactivity increases with increasing meat arc length. The results of the fuel meat volume computations for all 24 plates are provided in Table 6.10-2 for maximum fuel arc length.
The midpoint radii of the fuel plates are treated as fixed quantities in the NCT models, and are computed based on nominal dimensions. However, the channel width is modeled at the maximum value of 0.088-in between all plates in most NCT fuel element models. To achieve this channel width between all fuel plates, the cladding is artificially reduced to a thickness of 0.011-in, or a total plate thickness of 0.042-in. In the most reactive NCT models, a channel width of 0.092-in is modeled between all fuel plates. This value represents the maximum local channel width (0.090-in) plus an additional 0.002-in. To achieve this channel width between all fuel plates, the cladding is artificially reduced to a thickness of 0.009-in, or a total plate thickness of 0.038-in. These plate thicknesses are impossible to achieve in actual practice because they are below the allowable minimum plate thickness of 0.048-in.
The U-235 gram density for each fuel plate is computed by dividing the U-235 mass by the total volume, or 785 g/556.4 cm3 = 1.41 g/cm3. The fuel itself is a mixture of UAlx and aluminum.
The density of this mixture for ATR fuel is proportional to the U-235 gram density, as shown in Table 6.2-2. Because ATR and MURR fuel are of the same type, this equation is also used to develop the MURR fuel matrix density. These data are perfectly linear, and a linear fit of the data is 2 = 0.87331 + 2.5357, where 2 is the total gram density of the mixture, and 1 is the gram density of the U-235 in the mixture. Therefore, using this equation, the total density of the fuel matrix is computed to be approximately 3.77 g/cm3.
From the fuel volumes, U-235 gram densities, and total mixture densities provided, the number densities for the fuel region may be computed. These number densities are provided in Table 6.10-3. The U-235 weight percent is modeled at the maximum value of 94%. Representative weight percents of 0.6% and 0.35% are utilized for U-234 and U-236, respectively, and the balance (5.05%) is modeled as U-238.
6.10.2.2 MIT Fuel Element Each MIT element contains up to 515 g U-235, enriched up to 94 wt.%. The weight percents of the remaining uranium isotopes are 1.2 wt.% U-234, 0.7 wt.% U-236, and 5.0-7.0 wt.% U-238.
Each fuel element contains 15 flat fuel plates, as shown in Figure 6.10-2 and Figure 6.10-4. The fuel meat is a mixture of uranium metal and aluminum, while the cladding and structural materials are an aluminum alloy.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                    Rev. 16, May 2021 The geometry of the fuel element is defined in Figure 6.10-2. Each fuel plate is nominally 0.08-in thick, with a thickness tolerance of +/-0.003-in. The fuel meat is nominally 0.03-in thick, and the cladding is nominally 0.025-in thick. The plate cladding material is aluminum. Fuel element side plates are fabricated of ASTM B 209, aluminum alloy 6061-T6. These fuel element side plates have a nominal thickness of 0.188-in. The channel width between the plates is 0.078 +/-
0.004-in (excluding the thermal grooves). This tolerance represents average and not localized channel width. For an actual fuel element, the channel width may exceed this tolerance in localized areas. The maximum local channel width is 0.090-in (excluding the thermal grooves).
The maximum and minimum active fuel lengths and maximum and minimum active fuel widths may be computed based on Figure 6.10-2:
Maximum active fuel length = (23.0+0.01)-2(0.125) = 22.76-in Minimum active fuel length = (23.0-0.01)-2(0.5) = 21.99-in Maximum active fuel width = 2.531 - 2(0.18) = 2.171-in Minimum active fuel width = 2.521 - 2(0.27) = 1.981-in.
The nominal active fuel length may be estimated as the average of the maximum and minimum values, or 22.375-in.
It is necessary to determine the number densities of the fuel meat, which are the same for all fuel plates. To determine the number densities of the fuel meat, it is first necessary to compute the volume of the fuel meat. The volume of the fuel meat for each plate is the maximum width of the meat (2.171-in) multiplied by the active fuel length (22.375-in) and meat thickness (0.03-in).
The active fuel length and meat thickness are modeled at nominal values in all final (i.e., non-parametric) fuel element models, and the use of these dimensions is justified in Section 6.10.4.1.2, HAC Single Package Configuration. It is demonstrated that reactivity increases with increasing meat width. The total meat volume is therefore (15)(0.03)(22.375)(2.171)(2.543) =
358.2 cm3.
The centerlines of the fuel plates are treated as fixed quantities in the NCT models, and are computed based on nominal dimensions. However, the channel width is modeled at 0.094-in between all plates in most NCT fuel element models. This modeled channel width includes half of the thermal groove depth on each cladding plate. The fuel plates have grooves a maximum of 0.012-in deep cut into the surface of the fuel plates to increase heat transfer. Because the grooves cover approximately half the surface area of the cladding, half of the groove depth (i.e.,
0.006-in) is removed from each cladding plate, so that a channel width of 0.082+2*0.006 =
0.094-in is modeled. To achieve this channel width between all fuel plates, the cladding is artificially reduced to a thickness of 0.017-in, or a total plate thickness of 0.064-in.
Additional NCT models are developed in which the channel width is modeled at 0.116-in. This value is based upon the local maximum channel width (0.090-in) with an additional margin of 0.002-in, and the full thermal groove depth (0.012-in) removed from each plate (0.090-in +
0.002-in + 2*0.012-in = 0.116-in). To achieve this channel width between all fuel plates, the cladding is artificially reduced to a thickness of 0.006-in, or a total plate thickness of 0.042-in.
The U-235 gram density for each fuel plate is computed by dividing the U-235 mass by the total volume, or 515 g/358.2 cm3 = 1.44 g/cm3. The fuel itself is a mixture of UAlx and aluminum.
The density of this mixture for ATR fuel is proportional to the U-235 gram density, as shown in 6-107
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                Rev. 16, May 2021 Table 6.2-2. Because ATR and MIT fuel are of the same type, this equation is also used to develop the MIT fuel matrix density. These data are perfectly linear, and a linear fit of the data is 2 = 0.87331 + 2.5357, where 2 is the total gram density of the mixture, and 1 is the gram density of the U-235 in the mixture. Therefore, using this equation, the total density of the fuel matrix is computed to be approximately 3.79 g/cm3.
From the fuel volumes, U-235 gram densities, and total mixture densities provided, the number densities for the fuel region may be computed. These number densities are provided in Table 6.10-4. The U-235 weight percent is modeled at the maximum value of 94%. Representative weight percents of 0.6% and 0.35% are utilized for U-234 and U-236, respectively, and the balance (5.05%) is modeled as U-238.
Table 6.10 MURR Fuel Volume Computation (maximum arc length)
Midpoint        Fuel Arc      Volume Plate  Radius (cm)          (cm)        (cm3) 1        7.0993          4.5034      13.9460 2        7.4295          4.7625      14.7484 3        7.7597          5.0216      15.5507 4        8.0899          5.2832      16.3608 5        8.4201          5.5423      17.1632 6        8.7503          5.8014      17.9655 7        9.0805          6.0604      18.7678 8        9.4107          6.3195      19.5701 9        9.7409          6.5786      20.3724 10        10.0711          6.8377      21.1747 11        10.4013          7.0968      21.9770 12        10.7315          7.3558      22.7793 13        11.0617          7.6149      23.5816 14        11.3919          7.8765      24.3918 15        11.7221          8.1356      25.1941 16        12.0523          8.3947      25.9964 17        12.3825          8.6538      26.7987 18        12.7127          8.9129      27.6011 19        13.0429          9.1719      28.4034 20        13.3731          9.4310      29.2057 21        13.7033          9.6901      30.0080 22        14.0335          9.9492      30.8103 23        14.3637        10.2083      31.6126 24        14.6939        10.4699      32.4228 Total                    556.4024 6-108
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                        Rev. 16, May 2021 Table 6.10 MURR Fuel Number Densities (maximum arc length)
Number Density Isotope    (atom/b-cm)
U-234      2.3171E-05 U-235      3.6147E-03 U-236      1.3402E-05 U-238      1.9174E-04 Al      5.0596E-02 Total    5.4439E-02 Table 6.10 MIT Fuel Number Densities (maximum fuel width)
Number Density Isotope    (atom/b-cm)
U-234      2.3613E-05 U-235      3.6835E-03 U-236      1.3657E-05 U-238      1.9539E-04 Al      5.0481E-02 Total    5.4398E-02 6-109
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                                                              Rev. 16, May 2021 PLATE NO. DIM  A REF  DIM B  REF  DIM C  +/-
MIN      0.01 1        2.770      1.643    1.993 2        2.990      1.745    2.095 3        3.030      1.847    2.197
(.05) 4        3.160      1.950    2.300 5        3.290      2.052    2.402 6        3.420      2.154    2.504 7        3.550      2.256    2.606                                    R5.86 3.16 8        3.680      2.358    2.708 9        3.810      2.460    2.810 PLATE 24 10        3.940      2.562    2.912                                                                          CHANNEL WIDTH 11        4.070      2.664    3.014                                                                                .08 TYP 12        4.200      2.766    3.116 (45&deg;)
13        4.330      2.868    3.218 14        4.460      2.971    3.321
                                                                                                                                  .08 15        4.590      3.073    3.423 PLATE 1 (PLATES NUMBERED 16        4.720      3.175    3.525          CONSECUTIVELY FROM                                                      .15 17        4.850      3.277    3.627            PLATE 1 TO PLATE 24) 18        4.980      3.379    3.729 19        5.110      3.481    3.831                        CHANNEL 2 20        5.240      3.583    3.933                                                                      PLATE THICKNESS
(.15) 21        5.370      3.685    4.035                                                                          .05 TYP 22        5.500      3.787    4.137 23        5.630      3.889    4.239 24        5.760      3.992    4.342 2X 1.12 MAX                4X 1.00 MAXIMUM FUEL 2X .375 MIN                    4X 1.00    CORE BOUNDRY MINIMUM FUEL 4X .235                                          CORE BOUNDRY 4X .175 A
C 4X .355 B
25.50 2X .115 MIN FUEL PLATE DETAIL Figure 6.10 MURR Fuel Element Dimensions 6-110
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                              Rev. 16, May 2021 Figure 6.10 MIT Fuel Element Dimensions 6-111
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                            Rev. 16, May 2021 Fuel Plate 24 Fuel Plate 1 Water Channel Fuel Meat Figure 6.10 MURR Fuel Element Model 6-112
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                            Rev. 16, May 2021 Fuel Plate 15 Fuel Plate 1 Water Channel Fuel Meat Figure 6.10 MIT Fuel Element Model 6-113
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 6.10.3 General Considerations 6.10.3.1 Model Configuration The packaging is modeled essentially the same as described in Section 6.3.1, Model Configuration, including the number of packages utilized in the NCT and HAC array cases. The only difference is the FHE is modeled explicitly, and the contents are different.
The MURR and MIT FHEs are modeled explicitly over the active fuel length. The FHEs are constructed of aluminum. Maximum dimensional tolerances are selected so that the FHEs are as large as possible, which results in the largest possible pitch expansion in the HAC models. For the MURR FHE, these dimensions are 2.00+0.06-in, 3.56+0.06-in, 1.85+0.06-in, and 22.5&deg;+2&deg; (see the packaging general arrangement drawings for dimension placement). For the MIT FHE, these dimensions are 1.62+0.06-in and 2.82+0.06-in (see the packaging general arrangement drawings for dimension placement). The wall thickness is 0.19 +/- 0.06-in for each FHE. The array cases are run with both minimum and maximum wall thickness to determine the most reactive condition. All of the figures in this chapter show minimum wall thickness models.
Each FHE is comprised of two pieces held together by ball lock pins. Under NCT, the two FHE halves do not separate.
In the NCT single package models, the inner tube, FHE, insulation, and outer tube are modeled explicitly, as shown in Figure 6.10-5 and Figure 6.10-6 for MURR and MIT, respectively. An axial view is shown in Figure 6.10-7. Note that the thin steel sheet that encases the insulation has been conservatively neglected (the steel sheet would absorb neutrons and lower the reactivity). Although negligible water ingress is expected during NCT, the inner cavity of the package is assumed to be flooded with water because the package lid does not contain a seal.
However, the region between the insulation and the outer tube will remain dry because water cannot enter this region. In the models, the fuel element is conservatively positioned at the radial center of the FHE to maximize neutron reflection. The package is reflected with 12-in of full-density water.
The neoprene along the sides of the FHEs is modeled in an approximate manner using a thickness of 1/8-in. In both cases, the neoprene is modeled continuously along two sides for simplicity, rather than modeling the neoprene in detail as narrow strips. Because it was determined in the ATR fuel criticality analysis that neoprene will reduce the reactivity due to parasitic absorption in chlorine, the neoprene is modeled without chlorine, and the density is reduced accordingly.
The HAC single package model is similar to the NCT single package model. Damage in the drop tests was shown to be negligible and concentrated at the ends of the package (See Section 2.12.1, Certification Tests on CTU-1). As the ends of the package are not modeled, this end damage does not affect the modeling. The various side drops resulted in only minor localized damage to the outer tube, and no observable bulk deformation of the package. Therefore, the minor damage observed will not impact the reactivity. The insulation is replaced with full-density water, and the region between the insulation and outer tube is also filled with full-density water (see Figure 6.10-8 and Figure 6.10-9 for the MURR and MIT model geometry, respectively). The treatment of the fuel enclosure is the same as the NCT single package models.
Cases are developed both with and without the neoprene.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 No MURR or MIT fuels were included in the drop tests. Therefore, the damage to the MURR or MIT fuel under HAC is not known precisely. To conservatively bound the potential fuel damage in the HAC models, the fuel plate pitch is allowed to expand uniformly until constrained by the FHE. In addition, the FHEs, which are composed of two halves pinned together, are assumed to separate in a manner that maximizes the space available for pitch expansion. For simplicity, the gap between the two halves is not modeled explicitly in the HAC models. This pitch expansion increases the moderation and the reactivity. In actuality, such a large uniform expansion of the fuel element pitch is not credible, and in the worst case scenario would be localized at one end of the fuel element. Drop tests performed with ATR fuel, which is similar to MURR and MIT fuel, showed no damage that would affect the criticality analysis [See Section 2.12.1, Certification Tests on CTU-1]. The modeled damage is intended to bound a damaged fuel element that is otherwise intact.
In the NCT array models, a 9x9x1 array is utilized. To increase the reactivity, fuel elements are pushed toward the center of the array. Because the fuel elements are transported in a thin (~0.01-in) plastic bag, this plastic bag is allowed to act as a boundary for partial moderation effects. The plastic bag is not modeled explicitly, because it is too thin to have an appreciable effect on the reactivity. Therefore, it is postulated that the fuel element channels may fill with full-density water, while the region between the fuel element and FHE fills with variable density water.
Different water densities inside and outside the FHE are also addressed. Axial movement of the fuel elements is not considered because axial movement would increase the effective active height of the system (i.e., if some fuel elements shift and others remain in place) and reduce the reactivity due to increased leakage. The presence of chlorine-free neoprene is also considered in the array cases.
In the HAC array models, a 5x5x1 array is utilized, although the moderation conditions considered are similar to the NCT array analysis. Cases in which the insulation is replaced with water are also investigated. The fuel elements are modeled at the maximum pitch, consistent with the most reactive single package models.
The detailed moderation assumptions for these cases are discussed more fully in Section 6.10.5, Evaluation of Package Arrays under Normal Conditions of Transport, and Section 6.10.6, Package Arrays under Hypothetical Accident Conditions.
6.10.3.2 Material Properties The fuel meat compositions are provided in Table 6.10-3 and Table 6.10-4 for MURR and MIT fuel, respectively. The material properties of the packaging materials are provided in Section 6.3.2, Material Properties. The aluminum of the FHE is modeled as pure with a density of 2.7 g/cm3.
6.10.3.3 Computer Codes and Cross-Section Libraries The computer code and cross section libraries utilized are provided in Section 6.3.3, Computer Codes and Cross-Section Libraries.
6.10.3.4 Demonstration of Maximum Reactivity The reactivities of the NCT and HAC single package cases are small, with ks < 0.6.
6-115
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 For the NCT array, a 9x9x1 array is utilized, while for the HAC array, a smaller 5x5x1 array is utilized. Because negligible packaging damage was observed in the drop tests, the package dimensions are the same between the NCT and HAC models. However, the fuel elements are modeled differently between the NCT and HAC models. In the NCT models, the fuel elements are modeled as intact, although with dimensions optimized to maximize the reactivity. In the HAC models, the fuel is assumed to be damaged, and the pitch is allowed to expand until constrained by the FHE. In the HAC cases, the pins connecting the two halves of the FHE are assumed to break, and the two halves are pushed apart to the maximum extent to maximize the available space for pitch expansion. The FHEs and fuel elements are pushed toward the center of the array.
In both NCT and HAC array cases, flooding with partial moderation is allowed in the fuel element itself, between the fuel element and the FHE, and between the FHE and the inner tube.
A number of different partial moderation scenarios are considered.
In the NCT array models, insulation is modeled between the inner and outer tubes. In the HAC array models, it is demonstrated that modeling the insulation is more reactive than replacing the insulation with variable density water. In both sets of models, chlorine-free neoprene that is attached to the FHE is modeled, although the effect on the reactivity is small. No models in which the neoprene is allowed to decompose and homogeneously mix with the water are developed, as this scenario is already implicitly included in the search for optimum reactivity using various water densities.
Tolerances of the packaging materials are selected to maximize the reactivity. Both maximum and minimum wall thicknesses for the FHE are modeled to determine the most reactive condition, although the effect on the reactivity of this parameter is not significant.
The MURR fuel is significantly more reactive than the MIT fuel in all scenarios, a difference in ks of 0.186 comparing the most reactive models. The most reactive case occurs for the HAC array (Case XN9), and results in a ks = 0.85881, which is below the USL of 0.9209. For this case, full-density water is modeled between the fuel plates and inside the FHE, 0.8 g/cm3 water is modeled between the FHE and inner tube, the FHE is modeled with a thick wall, and insulation is modeled.
When comparing the reactivities of the three fuel types (ATR, MURR, MIT), MURR is the most reactive, MIT is the least reactive.
6-116
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                        Rev. 16, May 2021 12-in water reflector 1-in insulation 6.03-in 1/8-in neoprene 7.9-in Figure 6.10 MURR NCT Single Package Model (planar view) 6-117
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                        Rev. 16, May 2021 1/8-in neoprene Figure 6.10 MIT NCT Single Package Model (planar view) 6-118
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                      Rev. 16, May 2021 24.0-in                                    22.375-in MURR                                            MIT Note that the ends of both the fuel element and package are conservatively treated simply as a water reflector.
Figure 6.10 MURR/MIT NCT Single Package Models (axial view) 6-119
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                              Rev. 16, May 2021 Insulation and void replaced Expanded pitch with water.
Figure 6.10 MURR HAC Single Package Model (planar view) 6-120
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                Rev. 16, May 2021 Insulation and void replaced Expanded pitch with water.
Figure 6.10 MIT HAC Single Package Model (planar view) 6-121
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 6.10.4 Single Package Evaluation 6.10.4.1 Single Package Configuration Prior to development of a single package model, a parametric analysis is performed to determine the impacts of various fuel element tolerances on the reactivity. In the criticality analysis for ATR fuel (see Section 6.4.1.2.1, Fuel Element Payload Parametric Evaluation), it was determined that reactivity was maximized by maximizing the arc length of the fuel meat and the channel thickness. Because ATR, MURR, and MIT fuel are all plate-type and utilize similar enrichments, it is expected that MURR and MIT fuel will also experience maximum reactivity with these parameters maximized. Therefore, the parametric analysis considers the effects of only the following parameters: fuel meat arc length/width, channel width, and active fuel length.
The base configuration for both MURR and MIT consists of plates with a nominal meat arc length/width, nominal active fuel length, and nominal channel width. The minimum, nominal, and maximum meat arc lengths for MURR fuel are provided in Table 6.10-5. The minimum meat arc lengths are obtained directly from Figure 6.10-1 (see dimension B). The maximum meat arc lengths are computed by subtracting twice the fuel-free width (2*0.115-in) from the maximum plate width (dimension C of Figure 6.10-1 + 0.010-in). The nominal value is computed as the average of the minimum and maximum values.
A total of 14 parametric models are developed (7 for each fuel type), as listed in the following table. The detailed model descriptions of the parametric cases are summarized in Table 6.10-6.
In each parametric case, the indicated parameter is modified in comparison with the base case.
In all parametric models, the fuel element is modeled in the center of an ATR FFSC with the inner tube flooded, and the insulation replaced with full density water. The FHEs are neglected for simplicity.
Case ID                                        Case Description XB1        Base MURR case XB2        Decrease active fuel length to minimum value XB3        Increase active fuel length to maximum value XB4        Increase channel width to 0.088-in XB5        Decrease width of fuel meat to minimum value XB6        Increase width of fuel meat to maximum value XB7        Combine cases XB4 and XB6 Case ID                                        Case Description YB1        Base MIT case YB2        Decrease active fuel length to minimum value YB3        Increase active fuel length to maximum value YB4        Increase channel width to 0.094-in YB5        Decrease width of fuel meat to minimum value YB6        Increase width of fuel meat to maximum value YB7        Combine cases YB4 and YB6 The results of the parametric analysis are summarized in Table 6.10-7. Because the uncertainty in the calculation is ~0.001, a difference of at least 0.002 (2 milli-k, abbreviated mk) between the 6-122
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                    Rev. 16, May 2021 various cases is required in order to distinguish a real effect from statistical fluctuation. For both MURR and MIT fuel, the variation of the active fuel length has a negligible effect on the results.
Also, both MURR and MIT fuel show a positive reactivity increase when the fuel meat is widened and the channel width is increased. For MURR fuel, the increase is 23.5 mk (compare Case XB7 with Case XB1), and for MIT fuel, the increase is 8.8 mk (compare Case YB7 with Case YB1). This result is consistent with the results obtained in the ATR fuel analysis.
Therefore, in all subsequent NCT MURR and MIT fuel models, the fuel is modeled with nominal active fuel length and maximum fuel width. Also, in subsequent models, maximizing the channel width is achieved by either reducing the cladding thickness (if the fuel is undamaged) or increasing the plate pitch (if the fuel is damaged).
6.10.4.1.1      NCT Single Package Configuration The geometry of the NCT single package configuration is discussed in Section 6.10.3.1, Model Configuration. In the NCT single package models, the FHEs are modeled explicitly, and the neoprene is modeled in an approximate manner (see Figure 6.10-5 and Figure 6.10-6 for the NCT single package MURR and MIT models, respectively). The inner tube is flooded with full-density water. The fuel element geometry for both MURR and MIT is consistent with the most reactive fuel element model, including tolerances, as determined in the previous section.
Neoprene from the FHEs is modeled at the sides of the fuel element. Chlorine is conservatively removed from the neoprene because chlorine acts as a poison. The package is reflected with 12-in of water.
Results are provided in Table 6.10-8 for both MURR and MIT fuel. For MURR, Case XA1 is for a modeled channel width of 0.088-in, and Case XA2 is for a modeled channel width of 0.092-in. The channel width of 0.088-in represents the maximum average channel width, while the channel width of 0.092-in is the local maximum channel width of 0.090-in with an additional 0.002-in of margin. For MIT, Case YA1 is for a modeled channel width of 0.094-in, and Case YA2 is for a modeled channel width of 0.116-in. The channel width of 0.094-in represents the maximum average channel width (0.082-in) plus half of the thermal groove (0.006-in) on each cladding plate. The channel width of 0.116-in represents the local maximum channel width of 0.090-in plus the full thermal groove (0.012-in) on each cladding plate, plus an additional 0.002-in margin.
For both MURR and MIT, reactivity increases with increased channel width. The reactivity is low, with ks = 0.44807 for MURR and ks = 0.36978 for MIT. These results are below the USL of 0.9209.
6.10.4.1.2      HAC Single Package Configuration The geometry of the HAC single package configuration is discussed in Section 6.10.3.1, Model Configuration. In the HAC single package models, the FHEs are modeled explicitly, and the neoprene is modeled in an approximate manner (see Figure 6.10-8 and Figure 6.10-9 for the HAC single package MURR and MIT models, respectively). Chlorine is conservatively removed from the neoprene because chlorine acts as a poison. Eliminating the chlorine from the neoprene may be postulated to be a result of decomposition during a fire, although such a scenario is not credible.
The results are summarized in Table 6.10-9. In both the MURR and MIT models, the pitch is varied from the nominal value to the maximum value allowed by the FHE (Cases XC1 through 6-123
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 XC6 for MURR and YC1 through YC10 for MIT). For both fuel types, the reactivity increases as the plate pitch increases, reaching the maximum reactivity at the maximum pitch. For MURR, the maximum pitch is 0.167-in, which corresponds to a modeled channel spacing of 0.125-in.
For MIT, the maximum pitch is 0.240-in, which corresponds to a modeled channel spacing of 0.176-in. Neoprene is included in the variable pitch models. Note that the aluminum fuel element side plates are omitted from the MURR model for simplicity. In the MIT models, the aluminum fuel element side plates are allowed to stretch with the model for simplicity.
In Cases XC7 and YC11, the maximum-pitch MURR and MIT cases are repeated without neoprene. In both instances, the reactivity increases slightly when neoprene is modeled as water.
Because the fuel may be transported inside of a plastic bag, it is conservatively assumed that the water density inside of the FHE may vary independently of the water density inside of the fuel element. Note that additional surfaces are added to the MURR model to isolate the water between the fuel plates from the water inside the FHE (in Figure 6.10-8 these regions are combined). To maximize neutron reflection, full-density water is always modeled inside and outside the FHE, and the fuel element is centered laterally within the FHE.
In MURR Cases XC8 and XC9, Case XC7 is run with reduced water densities of 0.8 and 0.9 g/cm3 between the fuel plates, but maximum water density in all other regions of the model.
MIT Cases YC12 and YC13 are similar, except the Case YC11 is used as the base case. In both cases, reactivity drops as the water density is reduced between the fuel plates, indicating that the system is undermoderated.
The results are summarized in Table 6.10-9. Case XC7 is the most reactive MURR model, with ks = 0.54584, while Case YC11 is the most reactive MIT model, with ks = 0.43666. Both results are below the USL of 0.9209.
6.10.4.2 Single Package Results Following are the tabulated results for the single package cases. The most reactive configurations are listed in boldface.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                              Rev. 16, May 2021 Table 6.10 MURR Meat Arc Lengths Minimum    Nominal Maximum Plate (in)        (in)    (in) 1      1.643      1.708  1.773 2      1.745      1.810  1.875 3      1.847      1.912  1.977 4      1.950      2.015  2.080 5      2.052      2.117  2.182 6      2.154      2.219  2.284 7      2.256      2.321  2.386 8      2.358      2.423  2.488 9      2.460      2.525  2.590 10      2.562      2.627  2.692 11      2.664      2.729  2.794 12      2.766      2.831  2.896 13      2.868      2.933  2.998 14      2.971      3.036  3.101 15      3.073      3.138  3.203 16      3.175      3.240  3.305 17      3.277      3.342  3.407 18      3.379      3.444  3.509 19      3.481      3.546  3.611 20      3.583      3.648  3.713 21      3.685      3.750  3.815 22      3.787      3.852  3.917 23      3.889      3.954  4.019 24      3.992      4.057  4.122 6-125
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                      Rev. 16, May 2021 Table 6.10 Parametric Analysis Input Data MURR Parameter          XB1/XB4        XB2        XB3        XB5          XB6/XB7 Fuel width (in)            nominal    nominal    nominal  nominal-0.065  nominal+0.065 Meat thickness (in)          0.02        0.02      0.02      0.02            0.02 Active fuel height (in)        24      23.25      24.75        24              24 Channel (in)              0.08/0.088    0.08      0.08      0.08          0.08/0.088 Cladding (in)            0.015/0.011    0.015      0.015      0.015        0.015/0.011 Total plate (in)        0.050/0.042    0.050      0.050      0.050        0.050/0.042 Pitch (in)                  0.13        0.13      0.13      0.13            0.13 Meat volume (cm3)          544.13      527.13    561.14    531.86          556.40 U-235 mass (g)                785        785      785        785              785 U-235 den (g/cm3)            1.44        1.49      1.40      1.48            1.41 UAlx+Al den (g/cm3)          3.80        3.84      3.76      3.82            3.77 N-234 (atom/b-cm)        2.3694E-05  2.4458E-05 2.2976E-05  2.4241E-05      2.3171E-05 N-235 (atom/b-cm)        3.6962E-03  3.8154E-03 3.5842E-03  3.7815E-03      3.6147E-03 N-236 (atom/b-cm)        1.3704E-05  1.4146E-05 1.3289E-05  1.4020E-05      1.3402E-05 N-238 (atom/b-cm)        1.9607E-04  2.0239E-04 1.9012E-04  2.0059E-04      1.9174E-04 N-Al (atom/b-cm)        5.0460E-02  5.0262E-02 5.0646E-02  5.0319E-02      5.0596E-02 Total (atom/b-cm)        5.4390E-02  5.4319E-02 5.4457E-02  5.4339E-02      5.4439E-02 MIT Parameter          YB1/YB4        YB2        YB3        YB5          YB6/YB7 Fuel width (in)              2.076      2.076    2.076      1.981            2.171 Meat thickness (in)          0.03        0.03      0.03      0.03            0.03 Active fuel height (in)    22.375      21.99      22.76      22.375          22.375 Channel (in)            0.090/0.094    0.090      0.090      0.090        0.090/0.094 Cladding (in)            0.019/0.017    0.019      0.019      0.019        0.019/0.017 Total plate (in)        0.068/0.064    0.068      0.068      0.068        0.068/0.064 Pitch (in)                  0.158      0.158    0.158      0.158            0.158 Meat volume (cm3)          342.53      336.64    348.43    326.86          358.21 U-235 mass (g)                515        515      515        515              515 U-235 den (g/cm3)            1.503      1.530    1.478      1.576            1.438 UAlx+Al den (g/cm3)          3.85        3.87      3.83      3.91            3.79 N-234 (atom/b-cm)        2.4693E-05  2.5125E-05 2.4275E-05  2.5877E-05      2.3613E-05 N-235 (atom/b-cm)        3.8521E-03  3.9195E-03 3.7869E-03  4.0368E-03      3.6835E-03 N-236 (atom/b-cm)        1.4282E-05  1.4532E-05 1.4040E-05  1.4967E-05      1.3657E-05 N-238 (atom/b-cm)        2.0433E-04  2.0791E-04 2.0088E-04  2.1413E-04      1.9539E-04 N-Al (atom/b-cm)        5.0202E-02  5.0090E-02 5.0310E-02  4.9895E-02      5.0481E-02 Total (atom/b-cm)        5.4297E-02  5.4257E-02 5.4336E-02  5.4187E-02      5.4398E-02 6-126
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                              Rev. 16, May 2021 Table 6.10 Parametric Analysis Results ks        from Case ID        Filename          keff            (k+2)    XB1 (mk)
MURR XB1        HS_MURR2_P1        0.47068  0.00109  0.47286        --
XB2        HS_MURR2_P2        0.47199  0.00114  0.47427        1.4 XB3        HS_MURR2_P3        0.47075  0.00114  0.47303        0.2 XB4        HS_MURR2_P4        0.49257  0.00101  0.49459      21.7 XB5        HS_MURR2_P5        0.46808  0.00116  0.47040        -2.5 XB6        HS_MURR2_P6        0.47465  0.00097  0.47659        3.7 XB7        HS_MURR2_P7        0.49432  0.00102  0.49636      23.5 MIT ks        from Case ID        Filename          keff            (k+2)    YB1 (mk)
YB1          HS_MIT_P1        0.37801  0.00089  0.37979        --
YB2          HS_MIT_P2        0.37683  0.00093  0.37869        -1.1 YB3          HS_MIT_P3        0.37722  0.00091  0.37904        -0.8 YB4          HS_MIT_P4        0.38179  0.00095  0.38369        3.9 YB5          HS_MIT_P5        0.37018  0.00087  0.37192        -7.9 YB6          HS_MIT_P6        0.38064  0.00088  0.38240        2.6 YB7          HS_MIT_P7        0.38664  0.00097  0.38858        8.8 Table 6.10 NCT Single Package Results Moderator Density                            ks Case ID        Filename      (g/cm3)    keff              (k+2)
MURR XA1          NS_MURR          1.0    0.43268  0.00107    0.43482 XA2        NS_MURR2C          1.0    0.44597  0.00105    0.44807 MIT YA1          NS_MIT          1.0    0.33434  0.00086    0.33606 YA2          NS_MITC          1.0    0.36788  0.00095    0.36978 6-127
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                      Rev. 16, May 2021 Table 6.10 HAC Single Package Results Water Density Case                                            Between                          ks ID        Filename          Pitch (in)    Plates (g/cm3)    k            (k+2)
MURR XC1    HS_MURR2_NP00          0.130            1.0      0.48916 0.00107  0.49130 XC2    HS_MURR2_NP02          0.138            1.0      0.50506 0.00111  0.50728 XC3    HS_MURR2_NP04          0.146            1.0      0.51620 0.00116  0.51852 XC4    HS_MURR2_NP06          0.154            1.0      0.52285 0.00113  0.52511 XC5    HS_MURR2_NP08          0.161            1.0      0.53481 0.00104  0.53689 XC6    HS_MURR2_NP09          0.167            1.0      0.53887 0.00103  0.54093 XC7    HS_MURR2_P09          0.167            1.0      0.54374 0.00105  0.54584 XC8  HS_MURR2_P09_M080        0.167            0.8      0.47997 0.00111  0.48219 XC9  HS_MURR2_P09_M090        0.167            0.9      0.51244 0.00106  0.51456 MIT YC1      HS_MIT_NP158          0.158            1.0      0.37316 0.00090  0.37496 YC2      HS_MIT_NP16          0.160            1.0      0.37349 0.00095  0.37539 YC3      HS_MIT_NP17          0.170            1.0      0.38238 0.00088  0.38414 YC4      HS_MIT_NP18          0.180            1.0      0.38957 0.00098  0.39153 YC5      HS_MIT_NP19          0.190            1.0      0.39967 0.00105  0.40177 YC6      HS_MIT_NP20          0.200            1.0      0.40825 0.00095  0.41015 YC7      HS_MIT_NP21          0.210            1.0      0.41309 0.00104  0.41517 YC8      HS_MIT_NP22          0.220            1.0      0.41701 0.00100  0.41901 YC9      HS_MIT_NP23          0.230            1.0      0.42605 0.00093  0.42791 YC10      HS_MIT_NP24          0.240            1.0      0.43051 0.00105  0.43261 YC11      HS_MIT_P24          0.240            1.0      0.43474 0.00096  0.43666 YC12    HS_MIT_P24_M080        0.240            0.8      0.39439 0.00098  0.39635 YC13    HS_MIT_P24_M090        0.240            0.9      0.41226 0.00095  0.41416 6-128
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                    Rev. 16, May 2021 6.10.5 Evaluation of Package Arrays under Normal Conditions of Transport 6.10.5.1 NCT Array Configuration 6.10.5.1.1      MURR Fuel Element Models The NCT array model is a 9x9x1 array of the NCT single package model. Although an 8x8x1 array is of sufficient size to justify a CSI = 4.0, the larger 9x9x1 array is utilized simply for modeling convenience. Void is always present between the insulation and the outer tube, as this region is water-tight. The entire array is reflected with 12-in of full-density water.
The FHEs are pushed to the center of the array and rotated to minimize the distance between the fuel elements, see Figure 6.10-10. The modeled lateral shifting of the FHE inside of the tube is computed assuming the maximum inner diameter of the inner tube (5.814-in, see Section 6.3.1, Model Configuration) and minimum outer radius of the FHE (2.8-0.2 = 2.6-in, from the packaging general arrangement drawings), or 0.307-in. The fuel element is also modeled at the lateral top of the FHE to minimize the distance between the fuel elements.
Six calculational series are developed, as described below. Results are summarized in Table 6.10-10.
Series 1 (Cases XD1 through XD12): In Series 1, the water density is fixed at 1.0 g/cm3 between the fuel plates, and the water density inside and outside the FHE is modeled at the same density, which is allowed to vary between 0 and 1.0 g/cm3. This moderation condition simulates the partial moderation effect of assuming the plastic bag that surrounds the fuel element retains water. The neoprene (without chlorine) from the FHEs is modeled in an approximate manner.
The modeled channel width is 0.088-in. Also, the FHE is modeled with the minimum wall thickness.
As a point of interest, an additional case (Case XD12) is developed in which the fuel elements are centered in the cavity and not rotated, using the moderation assumptions of the most reactive case (Case XD7). The reactivity drops by 18.5 mk, which essentially represents the additional conservatism of pushing the fuel elements to the center of the array.
Series 2 (Cases XE1 through XE11): Series 2 is the same as Series 1, although the FHE neoprene is not modeled. The results in Table 6.10-10 indicate that the maximum reactivity occurs when chlorine-free neoprene is modeled (compare Cases XD7 and XE7), although the difference is within statistical fluctuation.
Series 3 (Cases XF1 through XF10): In Series 3, the water density inside the FHE is fixed at 1.0 g/cm3, while the water density outside the FHE is allowed to vary between 0 and 1.0 g/cm3. This moderation condition simulates the partial moderation effect of assuming the FHE retains water.
The maximum reactivity increases slightly compared to Series 1.
Series 4 (Cases XG1 through XG11): Series 4 is the same as Series 3, although the FHE is modeled with the maximum wall thickness. The reactivity increases slightly, although the difference is within statistical fluctuation.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                    Rev. 16, May 2021 Series 5 (Cases XH1 through XH11): Series 5 is the same as Series 3, although the density within the fuel plates is modeled at a reduced density of 0.9 g/cm3. The reactivity drops sharply as the water density between the plates is reduced.
Series 6 (Cases XI1 through XI11) is the same as Series 4, except the channel width is increased from 0.088-in to 0.092-in. The reactivity increases with increasing channel width, consistent with the single package models. Reactivity is at a maximum for Case XI5, with ks = 0.85643. In this case, the fuel elements are pushed to the center of the array, full-density water is modeled between the plates and inside the FHE, 0.4 g/cm3 water is modeled outside the FHE, chlorine-free neoprene is included, the FHE is modeled with maximum wall thickness, and the channel width is modeled at 0.092-in. The maximum result is below the USL of 0.9209.
6.10.5.1.2      MIT Fuel Element Models The NCT array model is a 9x9x1 array of the NCT single package model. Although an 8x8x1 array is of sufficient size to justify a CSI = 4.0, the larger 9x9x1 array is utilized simply for modeling convenience. Void is always present between the insulation and the outer tube, as this region is water-tight. The entire array is reflected with 12-in of full-density water.
The FHEs are pushed to the center of the array and rotated to minimize the distance between the fuel elements, see Figure 6.10-10. The modeled lateral shifting of the FHE inside of the tube is computed assuming the maximum inner diameter of the inner tube (5.814-in, see Section 6.3.1, Model Configuration) and minimum outer radius of the FHE (2.8-0.2 = 2.6-in, from the packaging general arrangement drawings), or 0.307-in.
In addition to the lateral shifting of the FHE within the tube, the MIT fuel element is free to move laterally within the FHE. To simplify the model geometry, rather than modeling each fuel element shifted within each FHE, the fuel elements are modeled in the center of the FHE, and the FHE is shifted toward the center of the array an additional 0.13-in (the approximate as-modeled distance between the fuel element and neoprene).
Six calculational series are developed, as described below. Results are summarized in Table 6.10-11.
Series 1 (Cases YD1 through YD12): In Series 1, the water density is fixed at 1.0 g/cm3 between the fuel plates, and the water density inside and outside the FHE is modeled at the same density, which is allowed to vary between 0 and 1.0 g/cm3. This moderation condition simulates the partial moderation effect of assuming the plastic bag that surrounds the fuel element retains water. The neoprene (without chlorine) from the FHE is modeled in an approximate manner.
The modeled channel width is 0.094-in. Also, the FHE is modeled with the minimum wall thickness.
As a point of interest, an additional case (Case YD12) is developed in which the fuel elements are centered in the cavity and not rotated, using the moderation assumptions of the most reactive case (Case YD7). The reactivity drops by 12.5 mk, which essentially represents the additional conservatism of pushing the fuel elements to the center of the array.
Series 2 (Cases YE1 through YE11): Series 2 is the same as Series 1, although the FHE neoprene is not modeled. Comparing Series 1 to Series 2, the reactivity is slightly higher when chlorine-free neoprene is modeled (compare Cases YD7 and YE7), although the difference is within statistical fluctuation.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 Series 3 (Cases YF1 through YF10): In Series 3, the water density inside the FHE is fixed at 1.0 g/cm3, while the water density outside the FHE is allowed to vary between 0 and 1.0 g/cm3. This moderation condition simulates the partial moderation effect of assuming the FHE retains water.
The maximum reactivity increases slightly compared to Series 1, although the effect is well within statistical fluctuation.
Series 4 (Cases YG1 through YG11): Series 4 is the same as Series 3, although the FHE is modeled with the maximum wall thickness. The reactivity decreases slightly, although the difference may be statistical fluctuation. Note that reactivity increased slightly with the thicker walled FHE in the MURR models.
Series 5 (Cases YH1 through YH11): Series 5 is the same as Series 3, although the density within the fuel plates is modeled at a reduced density of 0.9 g/cm3. The reactivity drops sharply as the water density between the plates is reduced.
Series 6 (Cases YI1 through YI11): Series 6 is the same as Series 3, although the modeled channel width is increased from 0.094-in to 0.116-in. Reactivity is at a maximum for Case YI6, with ks = 0.65658. In this case, the fuel elements are pushed to the center of the array, full-density water is modeled between the plates and inside the FHE, 0.5 g/cm3 water is modeled outside the FHE, chlorine-free neoprene is included, the FHE is modeled with minimum wall thickness, and the modeled channel width is 0.116-in. The maximum result is far below the USL of 0.9209.
6.10.5.2 NCT Array Results The results for the NCT array cases are provided in the following tables. The most reactive configuration in each series is listed in boldface.
6-131
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                Rev. 16, May 2021 Table 6.10 MURR NCT Array Results Water        Water Water      Density      Density Density      Outside    Between Case                              Inside FHE      FHE        Plates                        ks ID            Filename            (g/cm3)      (g/cm3)      (g/cm3)    keff            (k+2)
Series 1: Variable water density inside and outside FHE, with neoprene.
XD1      NA_MURR2_NW000              0            0          1.0  0.76937    0.00121  0.77179 XD2      NA_MURR2_NW010              0.1          0.1          1.0  0.79729    0.00123  0.79975 XD3      NA_MURR2_NW020              0.2          0.2          1.0  0.81129    0.00129  0.81387 XD4      NA_MURR2_NW030              0.3          0.3          1.0  0.82519    0.00129  0.82777 XD5      NA_MURR2_NW040              0.4          0.4          1.0  0.83449    0.00130  0.83709 XD6      NA_MURR2_NW050              0.5          0.5          1.0  0.83502    0.00123  0.83748 XD7      NA_MURR2_NW060              0.6          0.6          1.0  0.83801    0.00124  0.84049 XD8      NA_MURR2_NW070              0.7          0.7          1.0  0.83447    0.00111  0.83669 XD9      NA_MURR2_NW080              0.8          0.8          1.0  0.83185    0.00119  0.83423 XD10      NA_MURR2_NW090              0.9          0.9          1.0  0.82537    0.00123  0.82783 XD11      NA_MURR2_NW100              1.0          1.0          1.0  0.81935    0.00120  0.82175 XD12    NA_MURR2_NW060C              0.6          0.6          1.0  0.81957    0.00123  0.82203 Series 2: Repeat of Series 1 without neoprene XE1        NA_MURR2_W000              0            0          1.0  0.75717    0.00117  0.75951 XE2        NA_MURR2_W010              0.1          0.1          1.0  0.78680    0.00103  0.78886 XE3        NA_MURR2_W020              0.2          0.2          1.0  0.80910    0.00116  0.81142 XE4        NA_MURR2_W030              0.3          0.3          1.0  0.82154    0.00114  0.82382 XE5        NA_MURR2_W040              0.4          0.4          1.0  0.83148    0.00129  0.83406 XE6        NA_MURR2_W050              0.5          0.5          1.0  0.83479    0.00111  0.83701 XE7      NA_MURR2_W060              0.6          0.6          1.0  0.83681    0.00115  0.83911 XE8        NA_MURR2_W070              0.7          0.7          1.0  0.83504    0.00126  0.83756 XE9        NA_MURR2_W080              0.8          0.8          1.0  0.83138    0.00116  0.83370 XE10      NA_MURR2_W090              0.9          0.9          1.0  0.82487    0.00122  0.82731 XE11      NA_MURR2_W100              1.0          1.0          1.0  0.81734    0.00128  0.81990 Series 3: Variable water density outside FHE, with neoprene.
XF1      NA_MURR2_FNW000              1.0            0          1.0  0.83204    0.00135  0.83474 XF2      NA_MURR2_FNW010              1.0          0.1          1.0  0.83421    0.00118  0.83657 XF3      NA_MURR2_FNW020              1.0          0.2          1.0  0.84008    0.00131  0.84270 XF4      NA_MURR2_FNW030              1.0          0.3          1.0  0.84082    0.00132  0.84346 XF5      NA_MURR2_FNW040              1.0          0.4          1.0  0.84055    0.00120  0.84295 XF6      NA_MURR2_FNW050              1.0          0.5          1.0  0.83832    0.00116  0.84064 XF7      NA_MURR2_FNW060              1.0          0.6          1.0  0.83730    0.00118  0.83966 XF8      NA_MURR2_FNW070              1.0          0.7          1.0  0.83373    0.00130  0.83633 XF9      NA_MURR2_FNW080              1.0          0.8          1.0  0.83100    0.00124  0.83348 XF10    NA_MURR2_FNW090              1.0          0.9          1.0  0.82544    0.00129  0.82802 XD11      NA_MURR2_NW100              1.0          1.0          1.0  0.81935    0.00120  0.82175 (continued) 6-132
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                              Rev. 16, May 2021 Table 6.10 MURR NCT Array Results (concluded)
Water    Water      Water Density  Density    Density Inside  Outside  Between Case                                  FHE      FHE      Plates                          ks ID            Filename            (g/cm3)  (g/cm3)    (g/cm3)      keff          (k+2)
Series 4: Same as Series 3 but with maximum thickness FHE.
XG1      NA_MURR2_TFNW000              1.0        0        1.0      0.83659  0.00121  0.83901 XG2      NA_MURR2_TFNW010              1.0      0.1        1.0      0.83959  0.00114  0.84187 XG3      NA_MURR2_TFNW020              1.0      0.2        1.0      0.84116  0.00126  0.84368 XG4      NA_MURR2_TFNW030              1.0      0.3        1.0      0.84029  0.00128  0.84285 XG5      NA_MURR2_TFNW040              1.0      0.4        1.0      0.84340  0.00128  0.84596 XG6      NA_MURR2_TFNW050              1.0      0.5        1.0      0.83927  0.00116  0.84159 XG7      NA_MURR2_TFNW060              1.0      0.6        1.0      0.83816  0.00117  0.84050 XG8      NA_MURR2_TFNW070              1.0      0.7        1.0      0.83704  0.00131  0.83966 XG9      NA_MURR2_TFNW080              1.0      0.8        1.0      0.83199  0.00118  0.83435 XG10      NA_MURR2_TFNW090              1.0      0.9        1.0      0.82930  0.00116  0.83162 XG11      NA_MURR2_TFNW100              1.0      1.0        1.0      0.82461  0.00129  0.82719 3
Series 5: Same as Series 3 with 0.9 g/cm water between fuel plates.
XH1    NA_MURR2_M90FNW000            1.0        0        0.9      0.80160  0.00132  0.80424 XH2    NA_MURR2_M90FNW010            1.0      0.1        0.9      0.80747  0.00120  0.80987 XH3    NA_MURR2_M90FNW020            1.0      0.2        0.9      0.81288  0.00127  0.81542 XH4    NA_MURR2_M90FNW030              1.0      0.3        0.9      0.81512  0.00127  0.81766 XH5    NA_MURR2_M90FNW040            1.0      0.4        0.9      0.81504  0.00120  0.81744 XH6    NA_MURR2_M90FNW050            1.0      0.5        0.9      0.81382  0.00112  0.81606 XH7    NA_MURR2_M90FNW060            1.0      0.6        0.9      0.81369  0.00121  0.81611 XH8    NA_MURR2_M90FNW070            1.0      0.7        0.9      0.81165  0.00129  0.81423 XH9    NA_MURR2_M90FNW080            1.0      0.8        0.9      0.80950  0.00122  0.81194 XH10 NA_MURR2_M90FNW090                1.0      0.9        0.9      0.80311  0.00124  0.80559 XH11 NA_MURR2_M90FNW100                1.0      1.0        0.9      0.79735  0.00117  0.79969 Series 6: Same as Series 4 but with a modeled channel width of 0.092-in.
XI1      NA_MURR2_TFNW000C            1.0        0        1.0      0.84994  0.00110  0.85214 XI2      NA_MURR2_TFNW010C            1.0      0.1        1.0      0.85141  0.00120  0.85381 XI3      NA_MURR2_TFNW020C            1.0      0.2        1.0      0.85273  0.00124  0.85521 XI4      NA_MURR2_TFNW030C            1.0      0.3        1.0      0.85209  0.00124  0.85457 XI5    NA_MURR2_TFNW040C              1.0      0.4        1.0      0.85405  0.00119  0.85643 XI6      NA_MURR2_TFNW050C            1.0      0.5        1.0      0.84925  0.00127  0.85179 XI7      NA_MURR2_TFNW060C            1.0      0.6        1.0      0.84912  0.00124  0.85160 XI8      NA_MURR2_TFNW070C            1.0      0.7        1.0      0.84584  0.00115  0.84814 XI9      NA_MURR2_TFNW080C            1.0      0.8        1.0      0.84296  0.00127  0.84550 XI10    NA_MURR2_TFNW090C            1.0      0.9        1.0      0.83957  0.00115  0.84187 XI11    NA_MURR2_TFNW100C            1.0      1.0        1.0      0.83490  0.00123  0.83736 6-133
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                            Rev. 16, May 2021 Table 6.10 MIT NCT Array Results Water        Water      Water Density Density        Density        Between Case                        Inside FHE    Outside FHE        Plates                          ks ID        Filename            (g/cm3)      (g/cm3)        (g/cm3)    keff            (k+2)
Series 1: Variable water density and outside FHE, with neoprene YD1    NA_MIT_NW000              0            0              1.0    0.48041  0.00096  0.48233 YD2    NA_MIT_NW010            0.1          0.1              1.0    0.52918  0.00105  0.53128 YD3    NA_MIT_NW020            0.2          0.2              1.0    0.56301  0.00103  0.56507 YD4    NA_MIT_NW030            0.3          0.3              1.0    0.59062  0.00105  0.59272 YD5    NA_MIT_NW040            0.4          0.4              1.0    0.60722  0.00122  0.60966 YD6    NA_MIT_NW050            0.5          0.5              1.0    0.61575  0.00118  0.61811 YD7    NA_MIT_NW060            0.6          0.6              1.0    0.61989  0.00114  0.62217 YD8    NA_MIT_NW070            0.7          0.7              1.0    0.61723  0.00110  0.61943 YD9    NA_MIT_NW080            0.8          0.8              1.0    0.61618  0.00116  0.61850 YD10    NA_MIT_NW090            0.9          0.9              1.0    0.61352  0.00112  0.61576 YD11    NA_MIT_NW100            1.0          1.0              1.0    0.60885  0.00112  0.61109 YD12 NA_MIT_CNW060                0.6          0.6              1.0    0.60764  0.00103  0.60970 Series 2: Repeat of Series 1 without neoprene YE1      NA_MIT_W000              0            0              1.0    0.46154  0.00093  0.46340 YE2      NA_MIT_W010            0.1          0.1              1.0    0.51291  0.00095  0.51481 YE3      NA_MIT_W020            0.2          0.2              1.0    0.55394  0.00103  0.55600 YE4      NA_MIT_W030            0.3          0.3              1.0    0.58160  0.00113  0.58386 YE5      NA_MIT_W040            0.4          0.4              1.0    0.60184  0.00111  0.60406 YE6      NA_MIT_W050            0.5          0.5              1.0    0.61163  0.00119  0.61401 YE7      NA_MIT_W060            0.6          0.6              1.0    0.61746  0.00117  0.61980 YE8      NA_MIT_W070            0.7          0.7              1.0    0.61518  0.00116  0.61750 YE9      NA_MIT_W080            0.8          0.8              1.0    0.61215  0.00106  0.61427 YE10      NA_MIT_W090            0.9          0.9              1.0    0.61082  0.00111  0.61304 YE11      NA_MIT_W100            1.0          1.0              1.0    0.60324  0.00110  0.60544 Series 3: Variable water density outside FHE, with neoprene.
YF1    NA_MIT_FNW000            1.0            0              1.0    0.55417  0.00118  0.55653 YF2    NA_MIT_FNW010            1.0          0.1              1.0    0.57731  0.00104  0.57939 YF3    NA_MIT_FNW020            1.0          0.2              1.0    0.59825  0.00117  0.60059 YF4    NA_MIT_FNW030            1.0          0.3              1.0    0.60830  0.00119  0.61068 YF5    NA_MIT_FNW040            1.0          0.4              1.0    0.61581  0.00116  0.61813 YF6    NA_MIT_FNW050            1.0          0.5              1.0    0.61968  0.00107  0.62182 YF7    NA_MIT_FNW060            1.0          0.6              1.0    0.62059  0.00113  0.62285 YF8    NA_MIT_FNW070            1.0          0.7              1.0    0.62035  0.00110  0.62255 YF9    NA_MIT_FNW080            1.0          0.8              1.0    0.61650  0.00110  0.61870 YF10    NA_MIT_FNW090            1.0          0.9              1.0    0.61120  0.00105  0.61330 YD11    NA_MIT_NW100            1.0          1.0              1.0    0.60885  0.00112  0.61109 (continued) 6-134
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                            Rev. 16, May 2021 Table 6.10 MIT NCT Array Results (concluded)
Water    Water      Water Density    Density    Density Inside  Outside  Between Case                              FHE      FHE      Plates                          ks ID          Filename            (g/cm3)  (g/cm3)    (g/cm3)        keff          (k+2)
Series 4: Same as Series 3 but with maximum thickness FHE.
YG1      NA_MIT_TFNW000            1.0        0        1.0      0.55951  0.00106  0.56163 YG2      NA_MIT_TFNW010            1.0      0.1        1.0      0.58058  0.00105  0.58268 YG3      NA_MIT_TFNW020            1.0      0.2        1.0      0.59653  0.00105  0.59863 YG4      NA_MIT_TFNW030            1.0      0.3        1.0      0.60581  0.00118  0.60817 YG5      NA_MIT_TFNW040            1.0      0.4        1.0      0.61242  0.00110  0.61462 YG6      NA_MIT_TFNW050            1.0      0.5        1.0      0.61318  0.00104  0.61526 YG7      NA_MIT_TFNW060            1.0      0.6        1.0      0.61463  0.00120  0.61703 YG8      NA_MIT_TFNW070            1.0      0.7        1.0      0.61501  0.00111  0.61723 YG9      NA_MIT_TFNW080            1.0      0.8        1.0      0.61394  0.00114  0.61622 YG10      NA_MIT_TFNW090            1.0      0.9        1.0      0.60894  0.00113  0.61120 YG11      NA_MIT_TFNW100            1.0      1.0        1.0      0.60456  0.00120  0.60696 3
Series 5: Same as Series 3 with 0.9 g/cm water between fuel plates.
YH1    NA_MIT_M90FNW000            1.0        0        0.9      0.53177  0.00107  0.53391 YH2    NA_MIT_M90FNW010            1.0      0.1        0.9      0.55655  0.00108  0.55871 YH3    NA_MIT_M90FNW020            1.0      0.2        0.9      0.57776  0.00122  0.58020 YH4    NA_MIT_M90FNW030            1.0      0.3        0.9      0.59349  0.00102  0.59553 YH5    NA_MIT_M90FNW040            1.0      0.4        0.9      0.60205  0.00103  0.60411 YH6    NA_MIT_M90FNW050            1.0      0.5        0.9      0.60659  0.00102  0.60863 YH7    NA_MIT_M90FNW060            1.0      0.6        0.9      0.60651  0.00119  0.60889 YH8 NA_MIT_M90FNW070                1.0      0.7        0.9      0.60753  0.00121  0.60995 YH9    NA_MIT_M90FNW080            1.0      0.8        0.9      0.60615  0.00112  0.60839 YH10 NA_MIT_M90FNW090              1.0      0.9        0.9      0.60192  0.00100  0.60392 YH11 NA_MIT_M90FNW100              1.0      1.0        0.9      0.59396  0.00111  0.59618 Series 6: Same as Series 3 but with modeled channel width of 0.116-in.
YI1      NA_MIT_FNW000C            1.0        0        1.0      0.60247  0.00113  0.60473 YI2      NA_MIT_FNW010C            1.0      0.1        1.0      0.62391  0.00116  0.62623 YI3      NA_MIT_FNW020C            1.0      0.2        1.0      0.63710  0.00115  0.63940 YI4      NA_MIT_FNW030C            1.0      0.3        1.0      0.64617  0.00129  0.64875 YI5      NA_MIT_FNW040C            1.0      0.4        1.0      0.65160  0.00119  0.65398 YI6    NA_MIT_FNW050C            1.0      0.5        1.0      0.65414  0.00122  0.65658 YI7      NA_MIT_FNW060C            1.0      0.6        1.0      0.65181  0.00119  0.65419 YI8      NA_MIT_FNW070C            1.0      0.7        1.0      0.65016  0.00109  0.65234 YI9      NA_MIT_FNW080C            1.0      0.8        1.0      0.64541  0.00118  0.64777 YI10      NA_MIT_FNW090C            1.0      0.9        1.0      0.64029  0.00106  0.64241 YI11      NA_MIT_FNW100C            1.0      1.0        1.0      0.63436  0.00114  0.63664 6-135
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                        Rev. 16, May 2021 MURR Full view                    MURR Close-up MIT Full view                      MIT Close-up Figure 6.10 MURR/MIT NCT Array Geometry 6-136
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 6.10.6 Package Arrays under Hypothetical Accident Conditions 6.10.6.1 HAC Array Condition The HAC array model is a 5x5x1 array of packages. The primary difference comparing NCT to HAC is the modeled fuel damage, and separation of the FHE halves. Consistent with the HAC single package models, the two FHE halves are allowed to separate to the maximum possible extent, and the fuel element pitch is allowed to increase to the maximum possible value until constrained by the FHE. It is established in the HAC single package analysis that the reactivity is maximized with the maximum pitch, so all HAC array calculations utilize the maximum pitch.
The moderation conditions for the HAC array cases are largely the same as the NCT array moderation conditions, with the exception of the insulation region. In the HAC models, this region may be filled with variable density water. From the NCT array calculations, it was determined that the neoprene has a statistically insignificant effect on the reactivity, although the results showed a negligible increase. Therefore, neoprene is included in all HAC array models.
Also, it has also been established in the HAC single package and NCT array cases that reducing the water density between the fuel plates reduces the reactivity. Therefore, the water between the fuel plates is always modeled at full density.
Although it is not feasible in actual practice to push the FHEs to the center of the array if the two FHE halves are already pushed apart, both the MURR and MIT models are shifted by 0.307-in towards the center of the array, as determined in Section 6.10.5.1, NCT Array Configuration.
Note in Figure 6.10-11 that the FHEs for both MURR and MIT are sliced off in the corners because such a translation is not possible without interference, and the aluminum corners of the MIT element are also sliced off slightly for the same reason.
6.10.6.1.1      MURR Fuel Element Models Five calculational series are developed, as described below. Results are summarized in Table 6.10-12.
Series 1 (Cases XJ1 through XJ11): In Series 1, the water density inside and outside the FHE is modeled at the same density, which is allowed to vary between 0 and 1.0 g/cm3. This moderation condition simulates the partial moderation effect of assuming the plastic bag that surrounds the fuel element retains water. The region between the circular and square tubes is modeled as insulation/void, and the FHE is modeled with the minimum wall thickness.
Series 2 (Cases XK1 through XK11): In Series 2, the water density inside the FHE is fixed at 1.0 g/cm3, while the water density outside the FHE is allowed to vary between 0 and 1.0 g/cm3. This moderation condition simulates the partial moderation effect of assuming the FHE retains water.
The region between the circular and square tubes is modeled as insulation/void, and the FHE is modeled with a minimum wall thickness. The maximum reactivity increases slightly compared to Series 1, although the effect is well within statistical fluctuation.
An additional case (Case XK11) is developed in which the insulation is replaced with void for the most reactive Series 2 case (Case XK10). Comparing Cases XK10 and XK11, it is slightly more reactive to model the insulation, which is consistent with the trend in the ATR fuel analysis.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                    Rev. 16, May 2021 Series 3 (Cases XL1 through XL11): In Series 3, the outer insulation/void region is replaced with variable density water. There are now three regions that contain water: (1) between the circular and square tubes, (2) between FHE and circular tube, and (3) between fuel element and FHE. In this series, each of these regions is modeled with the same water density, which is allowed to vary between 0 and 1.0 g/cm3. Reactivity is significantly lower in Series 3 compared with either Series 1 or 2.
Series 4 (Cases XM1 through XM10): In Series 4, full-density water is modeled inside the FHE, while variable density water between 0 and 1.0 g/cm3 is modeled outside the FHE and between the inner and outer tubes. This series is less reactive than either Series 1 or 2.
Series 5 (Cases XN1 through XN11): Series 5 is a repeat of Series 2 except using a thick-walled FHE. The reactivity increases slightly when the thick-walled FHE is used.
Series 1, 2 and 5 result in similar reactivities within the statistical uncertainty of the method.
Case XN9 is the most reactive MURR case, with ks = 0.85881. In this case, the fuel elements are pushed to the center of the array, full-density water is modeled between the plates and inside the FHE, 0.8 g/cm3 water is modeled outside the FHE, insulation/void is modeled between the inner and outer tubes, chlorine-free neoprene is included, and the FHE is modeled with maximum wall thickness. The maximum result is below the USL of 0.9209.
6.10.6.1.2      MIT Fuel Element Models Five calculational series are developed, as described below. Results are summarized in Table 6.10-13.
Series 1 (Cases YJ1 through YJ11): In Series 1, the water density inside and outside the FHE is modeled at the same density, which is allowed to vary between 0 and 1.0 g/cm3. This moderation condition simulates the partial moderation effect of assuming the plastic bag that surrounds the fuel element retains water. The region between the circular and square tubes is modeled as insulation/void, and the FHE is modeled with the minimum wall thickness.
Series 2 (Cases YK1 through YK11): In Series 2, the water density inside the FHE is fixed at 1.0 g/cm3, while the water density outside the FHE is allowed to vary between 0 and 1.0 g/cm3. This moderation condition simulates the partial moderation effect of assuming the FHE retains water.
The region between the circular and square tubes is modeled as insulation/void, and the FHE is modeled with a minimum wall thickness. The maximum reactivity increases slightly compared to Series 1, although the effect is well within statistical fluctuation.
An additional case (Case YK11) is developed in which the insulation is replaced with void for the most reactive Series 2 case (Case YK9). Comparing Cases YK9 and YK11, it is slightly more reactive to model the insulation, which is consistent with the trend in the ATR fuel analysis.
Series 3 (Cases YL1 through YL11): In Series 3, the outer insulation/void region is replaced with variable density water. There are now three regions that contain water: (1) between the circular and square tubes, (2) between FHE and circular tube, and (3) between fuel element and FHE. In this series, each of these regions is modeled with the same water density, which is allowed to vary between 0 and 1.0 g/cm3. Reactivity is significantly lower in Series 3 compared with either Series 1 or 2.
6-138
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                    Rev. 16, May 2021 Series 4 (Cases YM1 through YM10): In Series 4, full-density water is modeled inside the FHE, while variable density water between 0 and 1.0 g/cm3 is modeled outside the FHE and between the inner and outer tubes. This series is less reactive than either Series 1 or 2.
Series 5 (Cases YN1 through YN11): Series 5 is a repeat of Series 2 except using a thick-walled FHE. The reactivity decreases slightly when the thick-walled FHE is used, although the decrease is within statistical fluctuation.
Series 1, 2 and 5 result in similar reactivities within the statistical uncertainty of the method.
Case YK9 is the most reactive MIT case, with ks = 0.67309. In this case, the fuel elements are pushed to the center of the array, full-density water is modeled between the plates and inside the FHE, 0.8 g/cm3 water is modeled outside the FHE, insulation/void is modeled between the inner and outer tubes, chlorine-free neoprene is included, and the FHE is modeled with minimum wall thickness. The maximum result is below the USL of 0.9209.
6.10.6.2 HAC Array Results Following are the tabulated results for the HAC array cases. The most reactive configuration in each series is listed in boldface.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                Rev. 16, May 2021 Table 6.10 MURR HAC Array Results Water        Water        Water Density      Density      Density Between        Inside      Outside Case                                  Tubes        FHE          FHE                              ks ID            Filename            (g/cm3)      (g/cm3)      (g/cm3)      keff            (k+2)
Series 1: Insulation modeled, full-density water between plates, variable density water as indicated.
XJ1      HA_MURR2_NW000              0            0            0      0.76355  0.00115    0.76585 XJ2      HA_MURR2_NW010              0            0.1          0.1      0.78430  0.00122    0.78674 XJ3      HA_MURR2_NW020              0            0.2          0.2      0.80290  0.00111    0.80512 XJ4      HA_MURR2_NW030              0            0.3          0.3      0.81874  0.00124    0.82122 XJ5      HA_MURR2_NW040              0            0.4          0.4      0.83311  0.00127    0.83565 XJ6      HA_MURR2_NW050              0            0.5          0.5      0.84140  0.00122    0.84384 XJ7      HA_MURR2_NW060              0            0.6          0.6      0.84544  0.00124    0.84792 XJ8      HA_MURR2_NW070              0            0.7          0.7      0.85035  0.00118    0.85271 XJ9      HA_MURR2_NW080              0            0.8          0.8      0.84998  0.00127    0.85252 XJ10      HA_MURR2_NW090              0            0.9          0.9      0.85379  0.00128    0.85635 XJ11      HA_MURR2_NW100              0            1.0          1.0      0.84975  0.00120    0.85215 Series 2: Insulation modeled, full-density water between plates and inside FHE, variable density water as indicated.
XK1      HA_MURR2_FNW000              0            1.0          0      0.83610  0.00115    0.83840 XK2      HA_MURR2_FNW010              0            1.0          0.1      0.84001  0.00125    0.84251 XK3      HA_MURR2_FNW020              0            1.0          0.2      0.84152  0.00115    0.84382 XK4      HA_MURR2_FNW030              0            1.0          0.3      0.84875  0.00130    0.85135 XK5      HA_MURR2_FNW040              0            1.0          0.4      0.84946  0.00127    0.85200 XK6      HA_MURR2_FNW050              0            1.0          0.5      0.84850  0.00119    0.85088 XK7      HA_MURR2_FNW060              0            1.0          0.6      0.85141  0.00118    0.85377 XK8      HA_MURR2_FNW070              0            1.0          0.7      0.85076  0.00117    0.85310 XK9      HA_MURR2_FNW080              0            1.0          0.8      0.85054  0.00127    0.85308 XK10      HA_MURR2_FNW090              0            1.0          0.9      0.85391  0.00125    0.85641 XJ11      HA_MURR2_NW100              0            1.0          1.0      0.84975    0.0012    0.85215 XK11    HA_MURR2_FNW090X              0            1.0          0.9      0.84922  0.00132    0.85186 Series 3: Insulation not modeled, variable density water as indicated.
XL1      HA_MURR2_ANW000              0            0            0      0.75710  0.00115    0.75940 XL2      HA_MURR2_ANW010              0.1          0.1          0.1      0.78773  0.00117    0.79007 XL3      HA_MURR2_ANW020              0.2          0.2          0.2      0.78883  0.00124    0.79131 XL4      HA_MURR2_ANW030              0.3          0.3          0.3      0.77894  0.00115    0.78124 XL5      HA_MURR2_ANW040              0.4          0.4          0.4      0.75950  0.00114    0.76178 XL6      HA_MURR2_ANW050              0.5          0.5          0.5      0.74010  0.00119    0.74248 XL7      HA_MURR2_ANW060              0.6          0.6          0.6      0.72381  0.00113    0.72607 XL8      HA_MURR2_ANW070              0.7          0.7          0.7      0.70323  0.00130    0.70583 XL9      HA_MURR2_ANW080              0.8          0.8          0.8      0.69154  0.00108    0.69370 XL10      HA_MURR2_ANW090              0.9          0.9          0.9      0.67881  0.00115    0.68111 XL11      HA_MURR2_ANW100              1.0          1.0          1.0      0.67207  0.00113    0.67433 (continued) 6-140
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                Rev. 16, May 2021 Table 6.10 MURR HAC Array Results (concluded)
Water      Water        Water Density    Density      Density Between      Inside      Outside Case                                Tubes        FHE          FHE                          ks ID            Filename            (g/cm3)    (g/cm3)      (g/cm3)    keff            (k+2)
Series 4: Insulation not modeled, variable density water as indicated.
XM1    HA_MURR2_IFNW000              0          1.0          0    0.83196  0.00121  0.83438 XM2      HA_MURR2_IFNW010            0.1          1.0          0.1    0.82347  0.00123  0.82593 XM3      HA_MURR2_IFNW020            0.2          1.0          0.2    0.80575  0.00127  0.80829 XM4      HA_MURR2_IFNW030            0.3          1.0          0.3    0.78652  0.00109  0.78870 XM5      HA_MURR2_IFNW040            0.4          1.0          0.4    0.76597  0.00108  0.76813 XM6      HA_MURR2_IFNW050            0.5          1.0          0.5    0.74360  0.00124  0.74608 XM7      HA_MURR2_IFNW060            0.6          1.0          0.6    0.72740  0.00119  0.72978 XM8      HA_MURR2_IFNW070            0.7          1.0          0.7    0.70952  0.00112  0.71176 XM9      HA_MURR2_IFNW080            0.8          1.0          0.8    0.69669  0.00115  0.69899 XM10      HA_MURR2_IFNW090            0.9          1.0          0.9    0.68144  0.00119  0.68382 XL11      HA_MURR2_ANW100            1.0          1.0          1.0    0.67207  0.00113  0.67433 Series 5: Repeat of Series 2 with thick-walled FHE.
XN1      HA_MURR2_TFNW000            0          1.0          0    0.83999  0.00136  0.84271 XN2      HA_MURR2_TFNW010            0          1.0          0.1    0.84169  0.00120  0.84409 XN3      HA_MURR2_TFNW020            0          1.0          0.2    0.84521  0.00115  0.84751 XN4      HA_MURR2_TFNW030            0          1.0          0.3    0.84875  0.00131  0.85137 XN5      HA_MURR2_TFNW040            0          1.0          0.4    0.84997  0.00117  0.85231 XN6      HA_MURR2_TFNW050            0          1.0          0.5    0.85368  0.00128  0.85624 XN7      HA_MURR2_TFNW060            0          1.0          0.6    0.85219  0.00115  0.85449 XN8      HA_MURR2_TFNW070            0          1.0          0.7    0.85204  0.00121  0.85446 XN9    HA_MURR2_TFNW080              0          1.0          0.8    0.85621  0.00130  0.85881 XN10      HA_MURR2_TFNW090            0          1.0          0.9    0.85319  0.00126  0.85571 XN11      HA_MURR2_TFNW100            0          1.0          1.0    0.85277  0.00121  0.85519 6-141
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 Table 6.10 MIT HAC Array Results Water        Water        Water Density      Density      Density Between        Inside      Outside Case                                Tubes        FHE          FHE                              ks ID          Filename            (g/cm3)      (g/cm3)      (g/cm3)        keff            (k+2)
Series 1: Insulation modeled, full-density water between plates, variable density water as indicated.
YJ1        HA_MIT_NW000              0            0            0        0.53667  0.00092  0.53851 YJ2        HA_MIT_NW010              0            0.1          0.1      0.56904  0.00111  0.57126 YJ3        HA_MIT_NW020              0            0.2          0.2      0.59837  0.00116  0.60069 YJ4        HA_MIT_NW030              0            0.3          0.3      0.62139  0.00122  0.62383 YJ5        HA_MIT_NW040              0            0.4          0.4      0.63737  0.00108  0.63953 YJ6        HA_MIT_NW050              0            0.5          0.5      0.65014  0.00109  0.65232 YJ7        HA_MIT_NW060              0            0.6          0.6      0.65850  0.00122  0.66094 YJ8        HA_MIT_NW070              0            0.7          0.7      0.66668  0.00115  0.66898 YJ9        HA_MIT_NW080              0            0.8          0.8      0.67043  0.00121  0.67285 YJ10        HA_MIT_NW090              0            0.9          0.9      0.67026  0.00112  0.67250 YJ11        HA_MIT_NW100              0            1.0          1.0      0.67058  0.00104  0.67266 Series 2: Insulation modeled, full-density water between plates and inside FHE, variable density water as indicated.
YK1        HA_MIT_FNW000              0            1.0          0        0.60486  0.00110  0.60706 YK2        HA_MIT_FNW010              0            1.0          0.1      0.62101  0.00117  0.62335 YK3        HA_MIT_FNW020              0            1.0          0.2      0.63436  0.00121  0.63678 YK4        HA_MIT_FNW030              0            1.0          0.3      0.64759  0.00106  0.64971 YK5        HA_MIT_FNW040              0            1.0          0.4      0.65646  0.00117  0.65880 YK6        HA_MIT_FNW050              0            1.0          0.5      0.66078  0.00117  0.66312 YK7        HA_MIT_FNW060              0            1.0          0.6      0.66656  0.00107  0.66870 YK8        HA_MIT_FNW070              0            1.0          0.7      0.67022  0.00114  0.67250 YK9        HA_MIT_FNW080              0            1.0          0.8      0.67105  0.00102  0.67309 YK10      HA_MIT_FNW090              0            1.0          0.9      0.66898  0.00113  0.67124 YJ11        HA_MIT_NW100              0            1.0          1.0      0.67058  0.00104  0.67266 YK11      HA_MIT_FNW080X              0            1.0          0.9      0.66684  0.00110  0.66904 Series 3: Insulation not modeled, variable density water as indicated.
YL1      HA_MIT_ANW000              0            0            0        0.53173  0.00103  0.53379 YL2      HA_MIT_ANW010              0.1          0.1          0.1      0.58121  0.00100  0.58321 YL3      HA_MIT_ANW020              0.2          0.2          0.2      0.59902  0.00119  0.60140 YL4        HA_MIT_ANW030              0.3          0.3          0.3      0.60054  0.00105  0.60264 YL5      HA_MIT_ANW040              0.4          0.4          0.4      0.59003  0.00116  0.59235 YL6      HA_MIT_ANW050              0.5          0.5          0.5      0.57811  0.00109  0.58029 YL7      HA_MIT_ANW060              0.6          0.6          0.6      0.56624  0.00114  0.56852 YL8      HA_MIT_ANW070              0.7          0.7          0.7      0.55438  0.00107  0.55652 YL9      HA_MIT_ANW080              0.8          0.8          0.8      0.54409  0.00114  0.54637 YL10      HA_MIT_ANW090              0.9          0.9          0.9      0.53935  0.00105  0.54145 YL11      HA_MIT_ANW100              1.0          1.0          1.0      0.53078    0.00104  0.53286 (continued) 6-142
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                              Rev. 16, May 2021 Table 6.10 MIT HAC Array Results (concluded)
Water        Water        Water Density      Density      Density Between        Inside      Outside Case                                Tubes        FHE          FHE                          ks ID          Filename            (g/cm3)      (g/cm3)      (g/cm3)    keff            (k+2)
Series 4: Insulation not modeled, variable density water as indicated.
YM1        HA_MIT_IFNW000              0            1.0          0    0.59996  0.00108  0.60212 YM2        HA_MIT_IFNW010            0.1          1.0          0.1  0.61992  0.00112  0.62216 YM3        HA_MIT_IFNW020            0.2          1.0          0.2  0.61899  0.00117  0.62133 YM4        HA_MIT_IFNW030            0.3          1.0          0.3  0.61130  0.00107  0.61344 YM5        HA_MIT_IFNW040            0.4          1.0          0.4  0.59725  0.00106  0.59937 YM6        HA_MIT_IFNW050            0.5          1.0          0.5  0.58253  0.00113  0.58479 YM7        HA_MIT_IFNW060            0.6          1.0          0.6  0.56935  0.00115  0.57165 YM8        HA_MIT_IFNW070            0.7          1.0          0.7  0.56002  0.00118  0.56238 YM9        HA_MIT_IFNW080            0.8          1.0          0.8  0.54870  0.00112  0.55094 YM10      HA_MIT_IFNW090            0.9          1.0          0.9  0.54119  0.00095  0.54309 YL11      HA_MIT_ANW100              1.0          1.0          1.0  0.53078  0.00104  0.53286 Series 5: Repeat of Series 2 with thick-walled FHE.
YN1      HA_MIT_TFNW000              0            1.0          0    0.61405  0.00116  0.61637 YN2      HA_MIT_TFNW010              0            1.0          0.1  0.62418  0.00114  0.62646 YN3      HA_MIT_TFNW020              0            1.0          0.2  0.63652  0.00110  0.63872 YN4      HA_MIT_TFNW030              0            1.0          0.3  0.64631  0.00101  0.64833 YN5      HA_MIT_TFNW040              0            1.0          0.4  0.65197  0.00108  0.65413 YN6      HA_MIT_TFNW050              0            1.0          0.5  0.65994  0.00114  0.66222 YN7      HA_MIT_TFNW060              0            1.0          0.6  0.66467  0.00118  0.66703 YN8      HA_MIT_TFNW070              0            1.0          0.7  0.66785  0.00120  0.67025 YN9      HA_MIT_TFNW080              0            1.0          0.8  0.66872  0.00123  0.67118 YN10      HA_MIT_TFNW090              0            1.0          0.9  0.66920  0.00111  0.67142 YN11      HA_MIT_TFNW100              0            1.0          1.0  0.66847  0.00122  0.67091 6-143
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                        Rev. 16, May 2021 MURR Full view                    MURR Close-up MIT Full view                      MIT Close-up Figure 6.10 MURR/MIT HAC Array Geometry 6-144
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 6.10.7 Fissile Material Packages for Air Transport See Section 6.7, which applies to all contents.
6.10.8 Benchmark Evaluations MURR and MIT fuel are both high-enriched aluminum plate-type fuel, similar to ATR fuel.
Therefore, the benchmarking evaluation performed for the ATR fuel in Section 6.8, Benchmark Evaluations, is applicable to the current analysis, and the USL is 0.9209. The Monte Carlo computer program MCNP5 v1.30 was utilized in the benchmark analysis. MCNP has been used extensively in criticality evaluations for several decades and is considered a standard in the industry.
Five parameters were selected for the benchmark evaluation: (1) energy of the average neutron lethargy causing fission (EALF), (2) U-235 number density, (3) channel width, (4) H/U-235 atom ratio, and (5) pitch. The range of applicability of these parameters for the benchmarks utilized is summarized in Table 6.8-2. In the following sections, the range of applicability of the benchmarks is compared with the MURR and MIT criticality analysis.
6.10.8.1 Energy of the Average neutron Lethargy causing Fission (EALF)
Range of Applicability, MURR models: All of the single package models and most of the NCT and HAC array models fall within the range of the applicability. The EALF of the most reactive MURR fuel element model (Case XN9) has an EALF of 9.26E-08 MeV, which is within the range of applicability. Models with significantly more void spaces or low water densities sometimes exceed the range of applicability (maximum EALF = 2.03E-07 MeV for Case XE1),
although these cases are not the most reactive. Therefore, the EALF of the most reactive models is acceptably within the range of applicability of the benchmarks.
Range of Applicability, MIT models: All of the single package models and most of the NCT and HAC array models fall within the range of the applicability. The EALF of the most reactive MIT fuel element model (Case YK9) has an EALF of 8.70E-08 MeV, which is within the range of applicability. Models with significantly more void spaces or low water densities sometimes exceed the range of applicability (maximum EALF = 3.30E-07 MeV for Case YE1), although these cases are not the most reactive. Therefore, the EALF of the most reactive models is acceptably within the range of applicability of the benchmarks.
6.10.8.2 U-235 Number Density The U-235 number density is 3.61E-03 atom/b-cm in the MURR models and 3.68E-03 atom/b-cm in the MIT models. These number densities are within the range of applicability.
6.10.8.3 Channel Width The maximum modeled NCT channel width is 0.092-in in the MURR models and 0.116-in in the MIT models. In the HAC models, in which the pitch is allowed to expand, the maximum channel width is 0.125-in in the MURR models and 0.176-in in the MIT models. All of these values exceed the maximum channel width of 0.078-in of the benchmark experiments. However, this parameter was artificially maximized in order to maximize model reactivity. As the channel width is directly related to system moderation, the acceptability of the EALF indicator 6-145
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                Rev. 16, May 2021 demonstrates that MCNP is performing acceptably for thermal conditions. Therefore, this parameter is considered to be acceptable.
6.10.8.4 H/U-235 Atom Ratio The H/U-235 atom ratio is used as the fourth trending parameter for the benchmark cases. The H/U-235 atom ratio is defined here as the ratio of hydrogen atoms to U-235 atoms in a unit cell.
This parameter is computed by the following equation:
NH*C/(NU235*M)
: where, NH is the hydrogen number density C is the channel width NU235 is the U-235 number density M is the fuel meat width Range of Applicability, MURR models: The H/U-235 atom ratio may be computed as:
NCT: 6.687E-02*0.088/(3.6147E-03*0.02) = 81.4 NCT: 6.687E-02*0.092/(3.6147E-03*0.02) = 85.1 HAC: 6.687E-02*0.125/(3.6147E-03*0.02) = 115.6 Therefore, H/U-235 of the MURR cases is acceptably within the range of applicability of the benchmarks.
Range of Applicability, MIT models: The H/U-235 atom ratio may be computed as:
NCT: 6.687E-02*0.094/(3.6835E-03*0.03) = 56.9 NCT: 6.687E-02*0.116/(3.6835E-03*0.03) = 70.2 HAC: 6.687E-02*0.176/(3.6835E-03*0.03) = 106.5 The minimum H/U-235 atom ratio of the benchmark models is 65.1. Therefore, this parameter is slightly outside the range of the benchmark experiments for the 0.094-in channel width NCT cases, although this parameter is in range for the more reactive 0.116-in channel width NCT cases. Therefore, this parameter is considered to be acceptable for the NCT cases. For the HAC cases, which bound the NCT cases, this parameter is acceptably within the range of applicability of the benchmarks.
6.10.8.5 Pitch The NCT pitch is fixed at 0.13-in in the MURR models and 0.16-in in the MIT models. In the HAC models, in which the pitch is allowed to expand, the maximum pitch is 0.167-in in the MURR models and 0.24-in in the MIT models. The maximum pitch of the benchmark models is 0.128-in, so the pitch in the models exceeds the range of the benchmarks, particularly for the HAC cases. However, this parameter was artificially maximized in order to maximize model reactivity. As the pitch is directly related to system moderation, the acceptability of the EALF indicator demonstrates that MCNP is performing acceptably for thermal conditions. Therefore, this parameter is considered to be acceptable.
6-146
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                            Rev. 16, May 2021 6.10.9 Sample Input Files A sample input file is provided for the most reactive MURR and MIT cases.
MURR Case XN9 (HA_MURR2_TFNW080)
MURR 999      0        -320:321:-322:323:-324:325                      imp:n=0 900      0        310 -311 312 -313 24 -25            fill=3      imp:n=1 901      2 -1.0 (311:-310:313:-312:-24:25) 320 -321 322 -323 324 -325              imp:n=1 c
c        Universe 1: MURR Fuel Element (infinitely long) c 10        10 5.4439E-02 52 16 -15                                    u=1 imp:n=1 $
plate 1 11        3 -2.7            (-52:53:16:15) 51 7 -8                  u=1 imp:n=1 12        10 5.4439E-02 401 -402 -406 -407                                u=1 imp:n=1 $
plate 2 13        3 -2.7            (-401:402:406:407) 400 -403 -404 -405 u=1 imp:n=1 14        10 5.4439E-02 411 -412 -416 -417                                u=1 imp:n=1 $
plate 3 15        3 -2.7            (-411:412:416:417) 410 -413 -414 -415 u=1 imp:n=1 16        10 5.4439E-02 421 -422 -426 -427                                u=1 imp:n=1 $
plate 4 17        3 -2.7            (-421:422:426:427) 420 -423 -424 -425 u=1 imp:n=1 18        10 5.4439E-02 431 -432 -436 -437                                u=1 imp:n=1 $
plate 5 19        3 -2.7            (-431:432:436:437) 430 -433 -434 -435 u=1 imp:n=1 20        10 5.4439E-02 441 -442 -446 -447                                u=1 imp:n=1 $
plate 6 21        3 -2.7            (-441:442:446:447) 440 -443 -444 -445 u=1 imp:n=1 22        10 5.4439E-02 451 -452 -456 -457                                u=1 imp:n=1 $
plate 7 23        3 -2.7            (-451:452:456:457) 450 -453 -454 -455 u=1 imp:n=1 24        10 5.4439E-02 461 -462 -466 -467                                u=1 imp:n=1 $
plate 8 25        3 -2.7            (-461:462:466:467) 460 -463 -464 -465 u=1 imp:n=1 26        10 5.4439E-02 471 -472 -476 -477                                u=1 imp:n=1 $
plate 9 27        3 -2.7            (-471:472:476:477) 470 -473 -474 -475 u=1 imp:n=1 28        10 5.4439E-02 481 -482 -486 -487                                u=1 imp:n=1 $
plate 10 29        3 -2.7            (-481:482:486:487) 480 -483 -484 -485 u=1 imp:n=1 30        10 5.4439E-02 491 -492 -496 -497                                u=1 imp:n=1 $
plate 11 31        3 -2.7            (-491:492:496:497) 490 -493 -494 -495 u=1 imp:n=1 32        10 5.4439E-02 501 -502 -506 -507                                u=1 imp:n=1 $
plate 12 33        3 -2.7            (-501:502:506:507) 500 -503 -504 -505 u=1 imp:n=1 34        10 5.4439E-02 511 -512 -516 -517                                u=1 imp:n=1 $
plate 13 35        3 -2.7            (-511:512:516:517) 510 -513 -514 -515 u=1 imp:n=1 36        10 5.4439E-02 521 -522 -526 -527                                u=1 imp:n=1 $
plate 14 37        3 -2.7            (-521:522:526:527) 520 -523 -524 -525 u=1 imp:n=1 38        10 5.4439E-02 531 -532 -536 -537                                u=1 imp:n=1 $
plate 15 6-147
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                Rev. 16, May 2021 39      3 -2.7        (-531:532:536:537) 530 -533 -534 -535 u=1 imp:n=1 40      10 5.4439E-02 541 -542 -546 -547                    u=1 imp:n=1 $
plate 16 41      3 -2.7        (-541:542:546:547) 540 -543 -544 -545 u=1 imp:n=1 42      10 5.4439E-02 551 -552 -556 -557                    u=1 imp:n=1 $
plate 17 43      3 -2.7        (-551:552:556:557) 550 -553 -554 -555 u=1 imp:n=1 44      10 5.4439E-02 561 -562 -566 -567                    u=1 imp:n=1 $
plate 18 45      3 -2.7        (-561:562:566:567) 560 -563 -564 -565 u=1 imp:n=1 46      10 5.4439E-02 571 -572 -576 -577                    u=1 imp:n=1 $
plate 19 47      3 -2.7        (-571:572:576:577) 570 -573 -574 -575 u=1 imp:n=1 48      10 5.4439E-02 581 -582 -586 -587                    u=1 imp:n=1 $
plate 20 49      3 -2.7        (-581:582:586:587) 580 -583 -584 -585 u=1 imp:n=1 50      10 5.4439E-02 591 -592 -596 -597                    u=1 imp:n=1 $
plate 21 51      3 -2.7        (-591:592:596:597) 590 -593 -594 -595 u=1 imp:n=1 52      10 5.4439E-02 601 -602 -606 -607                    u=1 imp:n=1 $
plate 22 53      3 -2.7        (-601:602:606:607) 600 -603 -604 -605 u=1 imp:n=1 54      10 5.4439E-02 611 -612 -616 -617                    u=1 imp:n=1 $
plate 23 55      3 -2.7        (-611:612:616:617) 610 -613 -614 -615 u=1 imp:n=1 56      10 5.4439E-02 621 -622 -626 -627                    u=1 imp:n=1 $
plate 24 57      3 -2.7        (-621:622:626:627) 620 -623 -624 -625 u=1 imp:n=1 150      2 -1.0  (-51:54:7:8)        (-400:403:404:405) (-410:413:414:415)
(-420:423:424:425) (-430:433:434:435) (-440:443:444:445)
(-450:453:454:455) (-460:463:464:465) (-470:473:474:475)
(-480:483:484:485) (-490:493:494:495) (-500:503:504:505)
(-510:513:514:515) (-520:523:524:525) (-530:533:534:535)
(-540:543:544:545) (-550:553:554:555) (-560:563:564:565)
(-570:573:574:575) (-580:583:584:585) (-590:593:594:595)
(-600:603:604:605) (-610:613:614:615) (-620:623:624:625) u=1 imp:n=1 c
c      Universe 19: MURR with FHE c
200    0        -232 -233 212 213 214 -234 fill=1(1) u=19 imp:n=1 201    5 -0.737 230 -210 212 214                  u=19 imp:n=1 $ right neoprene 202    5 -0.737 231 -211 213 214                  u=19 imp:n=1 $ left neoprene 203    2 -1.0    213 212 234                      u=19 imp:n=1 $ top water outside bag 204    2 -1.0    -230 232 214 212                u=19 imp:n=1 $ side water outside bag 205    2 -1.0    -231 233 214 213                u=19 imp:n=1 $ side water outside bag 206    3 -2.7    (210:211:-212:-213:-214) -220 -221 222 223 224 u=19 imp:n=1
$ FHE 207    2 -0.8    220:221:-222:-223:-224        u=19 imp:n=1 $ water c
c      Universe 20: MURR with pipe (center) c 6-148
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                              Rev. 16, May 2021 210    0            -200 fill=19              u=20 imp:n=1 211    4 -7.94      200 -201                  u=20 imp:n=1 $ pipe 212    6 -0.096    201 -203 250 -251 252 -253 u=20 imp:n=1 $ insulation 213    0            203 250 -251 252 -253      u=20 imp:n=1 $ insulation to tube 214    4 -7.94      -250:251:-252:253          u=20 imp:n=1 $ tube to inf c
c      Universe 21: MURR with pipe (down) c 220    0            -200 fill=19(2)            u=21 imp:n=1 221    4 -7.94      200 -201                  u=21 imp:n=1 $ pipe 222    6 -0.096    201 -203 250 -251 252 -253 u=21 imp:n=1 $ insulation 223    0            203 250 -251 252 -253      u=21 imp:n=1 $ insulation to tube 224    4 -7.94      -250:251:-252:253          u=21 imp:n=1 $ tube to inf c
c      Universe 22: MURR with pipe (up) c 230    0            -200 fill=19(3)            u=22 imp:n=1 231    4 -7.94      200 -201                  u=22 imp:n=1 $ pipe 232    6 -0.096    201 -203 250 -251 252 -253 u=22 imp:n=1 $ insulation 233    0            203 250 -251 252 -253      u=22 imp:n=1 $ insulation to tube 234    4 -7.94      -250:251:-252:253          u=22 imp:n=1 $ tube to inf c
c      Universe 23: MURR with pipe (right) c 240    0            -200 fill=19(4)            u=23 imp:n=1 241    4 -7.94      200 -201                  u=23 imp:n=1 $ pipe 242    6 -0.096    201 -203 250 -251 252 -253 u=23 imp:n=1 $ insulation 243    0            203 250 -251 252 -253      u=23 imp:n=1 $ insulation to tube 244    4 -7.94      -250:251:-252:253          u=23 imp:n=1 $ tube to inf c
c      Universe 24: MURR with pipe (left) c 250    0            -200 fill=19(5)            u=24 imp:n=1 251    4 -7.94      200 -201                  u=24 imp:n=1 $ pipe 252    6 -0.096    201 -203 250 -251 252 -253 u=24 imp:n=1 $ insulation 253    0            203 250 -251 252 -253      u=24 imp:n=1 $ insulation to tube 254    4 -7.94      -250:251:-252:253          u=24 imp:n=1 $ tube to inf c
c      Universe 25: MURR with pipe (up right) c 260    0            -200 fill=19(6)            u=25 imp:n=1 261    4 -7.94      200 -201                  u=25 imp:n=1 $ pipe 262    6 -0.096    201 -203 250 -251 252 -253 u=25 imp:n=1 $ insulation 263    0            203 250 -251 252 -253      u=25 imp:n=1 $ insulation to tube 264    4 -7.94      -250:251:-252:253          u=25 imp:n=1 $ tube to inf c
c      Universe 26: MURR with pipe (up left) c 270    0            -200 fill=19(7)            u=26 imp:n=1 271    4 -7.94      200 -201                  u=26 imp:n=1 $ pipe 272    6 -0.096    201 -203 250 -251 252 -253 u=26 imp:n=1 $ insulation 6-149
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                Rev. 16, May 2021 273    0          203 250 -251 252 -253        u=26 imp:n=1 $ insulation to tube 274    4 -7.94    -250:251:-252:253            u=26 imp:n=1 $ tube to inf c
c      Universe 27: MURR with pipe (down right) c 280    0          -200 fill=19(8)              u=27 imp:n=1 281    4 -7.94    200 -201                      u=27 imp:n=1 $ pipe 282    6 -0.096    201 -203 250 -251 252 -253    u=27 imp:n=1 $ insulation 283    0          203 250 -251 252 -253        u=27 imp:n=1 $ insulation to tube 284    4 -7.94    -250:251:-252:253            u=27 imp:n=1 $ tube to inf c
c      Universe 28: MURR with pipe (down left) c 290    0          -200 fill=19(9)              u=28 imp:n=1 291    4 -7.94    200 -201                      u=28 imp:n=1 $ pipe 292    6 -0.096    201 -203 250 -251 252 -253    u=28 imp:n=1 $ insulation 293    0          203 250 -251 252 -253        u=28 imp:n=1 $ insulation to tube 294    4 -7.94    -250:251:-252:253            u=28 imp:n=1 $ tube to inf c
c      Universe 3: Array of Packages c
300  0    -300 301 -302  303 imp:n=1 u=3 lat=1 fill=-2:2 -2:2 0:0 25 25 22 26  26 25 25 22 26  26 23 23 20 24  24 27 27 21 28  28 27 27 21 28  28 c 5      p 2.4142136 -1  0 -0.13275    $ right Al outer c 6      p -2.4142136 -1  0 -0.13275    $ left Al outer 7      p 2.4142136 -1 0  -1.09516  $ right Al inner 8      p -2.4142136 -1 0  -1.09516  $ left Al inner c 9      cz 6.858                      $ Al boundary c 10      cz 14.884                      $ Al boundary c
15      p 2.4142136 -1 0 -1.39997    $ plate meat boundary 16      p -2.4142136 -1 0 -1.39997    $ plate meat boundary c
24      pz -30.48                    $ bottom of fuel 25      pz 30.48                      $ top of fuel (24")
c 51      cz 7.0460 $ fuel plate 1 52      cz 7.0739 53      cz 7.1247 54      cz 7.1526 c
400 22    cz 7.3762 $ fuel plate 2 401 22    cz 7.4041 402 22    cz 7.4549 403 22    cz 7.4828 404 22    p 2.4142136 -1 0 -1.09516      $ right Al inner 405 22    p -2.4142136 -1 0 -1.09516    $ left Al inner 406 22    p 2.4142136 -1 0 -1.39997      $ plate meat boundary 407 22    p -2.4142136 -1 0 -1.39997    $ plate meat boundary 6-150
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                              Rev. 16, May 2021 c
410 23 cz 7.7064 $ fuel plate 3 411 23 cz 7.7343 412 23 cz 7.7851 413 23 cz 7.8130 414 23  p 2.4142136 -1 0 -1.09516    $ right Al inner 415 23  p -2.4142136 -1 0 -1.09516    $ left Al inner 416 23  p 2.4142136 -1 0 -1.39997    $ plate meat boundary 417 23  p -2.4142136 -1 0 -1.39997    $ plate meat boundary c
420 24 cz 8.0366 $ fuel plate 4 421 24 cz 8.0645 422 24 cz 8.1153 423 24 cz 8.1432 424 24  p 2.4142136 -1 0 -1.09516    $ right Al inner 425 24  p -2.4142136 -1 0 -1.09516    $ left Al inner 426 24  p 2.4142136 -1 0 -1.39997    $ plate meat boundary 427 24  p -2.4142136 -1 0 -1.39997    $ plate meat boundary c
430 25 cz 8.3668 $ fuel plate 5 431 25 cz 8.3947 432 25 cz 8.4455 433 25 cz 8.4734 434 25  p 2.4142136 -1 0 -1.09516    $ right Al inner 435 25  p -2.4142136 -1 0 -1.09516    $ left Al inner 436 25  p 2.4142136 -1 0 -1.39997    $ plate meat boundary 437 25  p -2.4142136 -1 0 -1.39997    $ plate meat boundary c
440 26 cz 8.6970 $ fuel plate 6 441 26 cz 8.7249 442 26 cz 8.7757 443 26 cz 8.8036 444 26  p 2.4142136 -1 0 -1.09516    $ right Al inner 445 26  p -2.4142136 -1 0 -1.09516    $ left Al inner 446 26  p 2.4142136 -1 0 -1.39997    $ plate meat boundary 447 26  p -2.4142136 -1 0 -1.39997    $ plate meat boundary c
450 27 cz 9.0272 $ fuel plate 7 451 27 cz 9.0551 452 27 cz 9.1059 453 27 cz 9.1338 454 27  p 2.4142136 -1 0 -1.09516    $ right Al inner 455 27  p -2.4142136 -1 0 -1.09516    $ left Al inner 456 27  p 2.4142136 -1 0 -1.39997    $ plate meat boundary 457 27  p -2.4142136 -1 0 -1.39997    $ plate meat boundary c
460 28 cz 9.3574 $ fuel plate 8 461 28 cz 9.3853 462 28 cz 9.4361 463 28 cz 9.4640 464 28  p 2.4142136 -1 0 -1.09516    $ right Al inner 465 28  p -2.4142136 -1 0 -1.09516    $ left Al inner 466 28  p 2.4142136 -1 0 -1.39997    $ plate meat boundary 467 28  p -2.4142136 -1 0 -1.39997    $ plate meat boundary c
470 29 cz 9.6876 $ fuel plate 9 471 29 cz 9.7155 6-151
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                Rev. 16, May 2021 472 29 cz 9.7663 473 29 cz 9.7942 474 29  p 2.4142136    -1 0 -1.09516    $ right Al inner 475 29  p -2.4142136  -1 0 -1.09516    $ left Al inner 476 29  p 2.4142136    -1 0 -1.39997    $ plate meat boundary 477 29  p -2.4142136  -1 0 -1.39997    $ plate meat boundary c
480 30 cz 10.0178 $ fuel  plate 10 481 30 cz 10.0457 482 30 cz 10.0965 483 30 cz 10.1244 484 30  p 2.4142136 -1    0 -1.09516    $ right Al inner 485 30  p -2.4142136 -1  0 -1.09516    $ left Al inner 486 30  p 2.4142136 -1    0 -1.39997    $ plate meat boundary 487 30  p -2.4142136 -1  0 -1.39997    $ plate meat boundary c
490 31 cz 10.3480 $ fuel  plate 11 491 31 cz 10.3759 492 31 cz 10.4267 493 31 cz 10.4546 494 31  p 2.4142136 -1    0 -1.09516    $ right Al inner 495 31  p -2.4142136 -1  0 -1.09516    $ left Al inner 496 31  p 2.4142136 -1    0 -1.39997    $ plate meat boundary 497 31  p -2.4142136 -1  0 -1.39997    $ plate meat boundary c
500 32 cz 10.6782 $ fuel  plate 12 501 32 cz 10.7061 502 32 cz 10.7569 503 32 cz 10.7848 504 32  p 2.4142136 -1    0 -1.09516    $ right Al inner 505 32  p -2.4142136 -1  0 -1.09516    $ left Al inner 506 32  p 2.4142136 -1    0 -1.39997    $ plate meat boundary 507 32  p -2.4142136 -1  0 -1.39997    $ plate meat boundary c
510 33 cz 11.0084 $ fuel  plate 13 511 33 cz 11.0363 512 33 cz 11.0871 513 33 cz 11.1150 514 33  p 2.4142136 -1    0 -1.09516    $ right Al inner 515 33  p -2.4142136 -1  0 -1.09516    $ left Al inner 516 33  p 2.4142136 -1    0 -1.39997    $ plate meat boundary 517 33  p -2.4142136 -1  0 -1.39997    $ plate meat boundary c
520 34 cz 11.3386 $ fuel  plate 14 521 34 cz 11.3665 522 34 cz 11.4173 523 34 cz 11.4452 524 34  p 2.4142136 -1    0 -1.09516    $ right Al inner 525 34  p -2.4142136 -1  0 -1.09516    $ left Al inner 526 34  p 2.4142136 -1    0 -1.39997    $ plate meat boundary 527 34  p -2.4142136 -1  0 -1.39997    $ plate meat boundary c
530 35 cz 11.6688 $ fuel plate 15 531 35 cz 11.6967 532 35 cz 11.7475 533 35 cz 11.7754 534 35  p 2.4142136 -1 0 -1.09516      $ right Al inner 6-152
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                Rev. 16, May 2021 535 35  p -2.4142136 -1 0 -1.09516      $ left Al inner 536 35  p 2.4142136 -1 0 -1.39997      $ plate meat boundary 537 35  p -2.4142136 -1 0 -1.39997      $ plate meat boundary c
540 36 cz 11.9990 $ fuel  plate 16 541 36 cz 12.0269 542 36 cz 12.0777 543 36 cz 12.1056 544 36  p 2.4142136 -1    0 -1.09516    $ right Al inner 545 36  p -2.4142136 -1  0 -1.09516    $ left Al inner 546 36  p 2.4142136 -1    0 -1.39997    $ plate meat boundary 547 36  p -2.4142136 -1  0 -1.39997    $ plate meat boundary c
550 37 cz 12.3292 $ fuel  plate 17 551 37 cz 12.3571 552 37 cz 12.4079 553 37 cz 12.4358 554 37  p 2.4142136 -1    0 -1.09516    $ right Al inner 555 37  p -2.4142136 -1  0 -1.09516    $ left Al inner 556 37  p 2.4142136 -1    0 -1.39997    $ plate meat boundary 557 37  p -2.4142136 -1  0 -1.39997    $ plate meat boundary c
560 38 cz 12.6594 $ fuel  plate 18 561 38 cz 12.6873 562 38 cz 12.7381 563 38 cz 12.7660 564 38  p 2.4142136 -1    0 -1.09516    $ right Al inner 565 38  p -2.4142136 -1  0 -1.09516    $ left Al inner 566 38  p 2.4142136 -1    0 -1.39997    $ plate meat boundary 567 38  p -2.4142136 -1  0 -1.39997    $ plate meat boundary c
570 39 cz 12.9896 $ fuel  plate 19 571 39 cz 13.0175 572 39 cz 13.0683 573 39 cz 13.0962 574 39  p 2.4142136 -1    0 -1.09516    $ right Al inner 575 39  p -2.4142136 -1  0 -1.09516    $ left Al inner 576 39  p 2.4142136 -1    0 -1.39997    $ plate meat boundary 577 39  p -2.4142136 -1  0 -1.39997    $ plate meat boundary c
580 40 cz 13.3198 $ fuel  plate 20 581 40 cz 13.3477 582 40 cz 13.3985 583 40 cz 13.4264 584 40  p 2.4142136 -1    0 -1.09516    $ right Al inner 585 40  p -2.4142136 -1  0 -1.09516    $ left Al inner 586 40  p 2.4142136 -1    0 -1.39997    $ plate meat boundary 587 40  p -2.4142136 -1  0 -1.39997    $ plate meat boundary c
590 41 cz 13.6500 $ fuel  plate 21 591 41 cz 13.6779 592 41 cz 13.7287 593 41 cz 13.7566 594 41  p 2.4142136 -1    0 -1.09516    $ right Al inner 595 41  p -2.4142136 -1  0 -1.09516    $ left Al inner 596 41  p 2.4142136 -1    0 -1.39997    $ plate meat boundary 597 41  p -2.4142136 -1  0 -1.39997    $ plate meat boundary 6-153
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                Rev. 16, May 2021 c
600 42 cz 13.9802 $ fuel  plate 22 601 42 cz 14.0081 602 42 cz 14.0589 603 42 cz 14.0868 604 42  p 2.4142136 -1    0 -1.09516    $ right Al inner 605 42  p -2.4142136 -1  0 -1.09516    $ left Al inner 606 42  p 2.4142136 -1    0 -1.39997    $ plate meat boundary 607 42  p -2.4142136 -1  0 -1.39997    $ plate meat boundary c
610 43 cz 14.3104 $ fuel  plate 23 611 43 cz 14.3383 612 43 cz 14.3891 613 43 cz 14.4170 614 43  p 2.4142136 -1    0 -1.09516    $ right Al inner 615 43  p -2.4142136 -1  0 -1.09516    $ left Al inner 616 43  p 2.4142136 -1    0 -1.39997    $ plate meat boundary 617 43  p -2.4142136 -1  0 -1.39997    $ plate meat boundary c
620 44 cz 14.6406 $ fuel  plate 24 621 44 cz 14.6685 622 44 cz 14.7193 623 44 cz 14.7472 624 44  p 2.4142136 -1    0 -1.09516    $ right Al inner 625 44  p -2.4142136 -1  0 -1.09516    $ left Al inner 626 44  p 2.4142136 -1    0 -1.39997    $ plate meat boundary 627 44  p -2.4142136 -1  0 -1.39997    $ plate meat boundary c
200    cz 7.3838 $ IR pipe 201    cz 7.6581 $ OR pipe c 202    cz 38.1    $ 12" water 203    cz 10.1981 $ 1" insulation c
210 50    p 2.194300 -1 0 11.6987    $ right lower inner 211 51    p -2.194300 -1 0 11.6987  $ left lower inner 212 50    p -0.455726 -1 0 -5.7501  $ right upper inner 213 51    p 0.455726 -1 0 -5.7501    $ left upper inner 214      py -5.6175                $ bottom inner 220 50    p 2.194300 -1 0 13.2300    $ right lower outer 221 51    p -2.194300 -1 0 13.2300  $ left lower outer 222 50    p -0.455726 -1 0 -6.4479  $ right upper outer 223 51    p 0.455726 -1 0 -6.4479    $ left upper outer 224      py -6.2525                $ bottom outer 230 50    p 2.194300 -1 0 10.9331    $ right neoprene 231 51    p -2.194300 -1 0 10.9331  $ left neoprene 232      p 3.1993      -1 0 13.2244 $ right plastic bag 233      p -3.1993    -1 0 13.2244 $ left plastic bag 234      c/z 0 -10.065 14.8        $ top of plastic bag c
250    px  -9.6032 $ square tube 251    px    9.6032 252    py  -9.6032 253    py    9.6032 c
300    px 10.033 $ lattice surfaces/sq. tube 301    px -10.033 302    py 10.033 6-154
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                Rev. 16, May 2021 303      py -10.033 310      px -50.165 $ 5x5 bounds 311      px  50.165 312      py -50.165 313      py  50.165 320      px -80.645 $ outer bounds 321      px  80.645 322      py -80.645 323      py  80.645 324      pz -60.96 325      pz  60.96 m2      1001.62c 2            $ water 8016.62c 1 mt2    lwtr.60t m3      13027.62c 1          $ Al m4      6000.66c  -0.08      $ SS-304 14000.60c -1.0 15031.66c -0.045 24000.50c -19.0 25055.62c -2.0 26000.55c -68.375 28000.50c -9.5 m5      1001.62c -0.056920    $ neoprene (no Cl) 6000.66c -0.542646 c        17000.66c -0.400434 m6      13027.62c -26.5      $ insulation material 14000.60c -23.4 8016.62c  -50.2 m10    92234.69c 2.3171E-05 92235.69c 3.6147E-03 92236.69c 1.3402E-05 92238.69c 1.9174E-04 13027.62c 5.0596E-02 c          total 5.4439E-02 c
*tr1    0 -12.25 0                        $ base to center
*tr2    0 -0.7798 0 180 90 90 90 180 90 $ down
*tr3    0 0.7798 0                        $ up
*tr4    0.7798 0 0 90 180 90 0 90 90      $ right
*tr5    -0.7798 0 0 90 0 90 180 90 90      $ left
*tr6    0.5514  0.5514 0 45 135 90 45 45 90        $ up/right
*tr7    -0.5514  0.5514 0 45 45 90 135 45 90        $ up/left
*tr8    0.5514 -0.5514 0 135 135 90 45 135 90      $ down/right
*tr9    -0.5514 -0.5514 0 135 45 90 135 135 90      $ down/left tr22 0 0.095 0 $ plate 2 tr23 0 0.190 0 $ plate 3 tr24 0 0.285 0 $ plate 4 tr25 0 0.380 0 $ plate 5 tr26 0 0.475 0 $ plate 6 tr27 0 0.570 0 $ plate 7 tr28 0 0.665 0 $ plate 8 tr29 0 0.760 0 $ plate 9 tr30 0 0.855 0 $ plate 10 tr31 0 0.950 0 $ plate 11 tr32 0 1.045 0 $ plate 12 tr33 0 1.140 0 $ plate 13 6-155
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                Rev. 16, May 2021 tr34 0 1.235 0 $ plate 14 tr35 0 1.330 0 $ plate 15 tr36 0 1.425 0 $ plate 16 tr37 0 1.520 0 $ plate 17 tr38 0 1.615 0 $ plate 18 tr39 0 1.710 0 $ plate 19 tr40 0 1.805 0 $ plate 20 tr41 0 1.900 0 $ plate 21 tr42 0 1.995 0 $ plate 22 tr43 0 2.090 0 $ plate 23 tr44 0 2.185 0 $ plate 24 tr50  0.7798 0 0 $ shift FHE right tr51 -0.7798 0 0 $ shift FHE left c
mode  n kcode 2500 1.0 50 250 sdef    x=d1 y=d2 z=d3 si1    -50 50 sp1    0 1 si2    -50 50 sp2    0 1 si3    -31 31 sp3    0 1 MIT Case YK9 (HA_MIT_FNW080)
MIT 999    0      -320:321:-322:323:-324:325                imp:n=0 900    0      310 -311 312 -313 24 -25    fill=3      imp:n=1 901    2 -1.0 (311:-310:313:-312:-24:25) 320 -321 322 -323 324 -325    imp:n=1 c
c      Universe 1: MIT Fuel Element (infinitely long) c 10      3 -2.7        10 -11 50 -124          u=1 imp:n=1 $  right Al piece 11      3 -2.7        13 -12 50 -124          u=1 imp:n=1 $  left Al piece c 12      2 -1.0        12 -10 18 -50            u=1 imp:n=1 20      10 5.4398E-02  40 -41 70 -90            u=1 imp:n=1 $  plate 1 21      3 -2.7        12 -10 50 -110 #20      u=1 imp:n=1 22      2 -1.0        12 -10 110 -51          u=1 imp:n=1 30      10 5.4398E-02  40 -41 71 -91            u=1 imp:n=1 $  plate 2 31      3 -2.7        12 -10 51 -111 #30      u=1 imp:n=1 32      2 -1.0        12 -10 111 -52          u=1 imp:n=1 40      10 5.4398E-02  40 -41 72 -92            u=1 imp:n=1 $  plate 3 41      3 -2.7        12 -10 52 -112 #40      u=1 imp:n=1 42      2 -1.0        12 -10 112 -53          u=1 imp:n=1 50      10 5.4398E-02  40 -41 73 -93            u=1 imp:n=1 $  plate 4 51      3 -2.7        12 -10 53 -113 #50      u=1 imp:n=1 52      2 -1.0        12 -10 113 -54          u=1 imp:n=1 60      10 5.4398E-02  40 -41 74 -94            u=1 imp:n=1 $  plate 5 61      3 -2.7        12 -10 54 -114 #60      u=1 imp:n=1 62      2 -1.0        12 -10 114 -55          u=1 imp:n=1 70      10 5.4398E-02  40 -41 75 -95            u=1 imp:n=1 $  plate 6 71      3 -2.7        12 -10 55 -115 #70      u=1 imp:n=1 72      2 -1.0        12 -10 115 -56          u=1 imp:n=1 80      10 5.4398E-02  40 -41 76 -96            u=1 imp:n=1 $  plate 7 81      3 -2.7        12 -10 56 -116 #80      u=1 imp:n=1 6-156
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                Rev. 16, May 2021 82      2 -1.0        12 -10 116 -57          u=1 imp:n=1 90      10 5.4398E-02 40 -41 77 -97            u=1 imp:n=1 $ plate 8 91      3 -2.7        12 -10 57 -117 #90      u=1 imp:n=1 92      2 -1.0        12 -10 117 -58          u=1 imp:n=1 100      10 5.4398E-02 40 -41 78 -98            u=1 imp:n=1 $ plate 9 101      3 -2.7        12 -10 58 -118 #100      u=1 imp:n=1 102      2 -1.0        12 -10 118 -59          u=1 imp:n=1 110      10 5.4398E-02 40 -41 79 -99            u=1 imp:n=1 $ plate 10 111      3 -2.7        12 -10 59 -119 #110      u=1 imp:n=1 112      2 -1.0        12 -10 119 -60          u=1 imp:n=1 120      10 5.4398E-02 40 -41 80 -100            u=1 imp:n=1 $ plate 11 121      3 -2.7        12 -10 60 -120 #120      u=1 imp:n=1 122      2 -1.0        12 -10 120 -61          u=1 imp:n=1 130      10 5.4398E-02 40 -41 81 -101            u=1 imp:n=1 $ plate 12 131      3 -2.7        12 -10 61 -121 #130      u=1 imp:n=1 132      2 -1.0        12 -10 121 -62          u=1 imp:n=1 140      10 5.4398E-02 40 -41 82 -102            u=1 imp:n=1 $ plate 13 141      3 -2.7        12 -10 62 -122 #140      u=1 imp:n=1 142      2 -1.0        12 -10 122 -63          u=1 imp:n=1 150      10 5.4398E-02 40 -41 83 -103            u=1 imp:n=1 $ plate 14 151      3 -2.7        12 -10 63 -123 #150      u=1 imp:n=1 152      2 -1.0        12 -10 123 -64          u=1 imp:n=1 160      10 5.4398E-02 40 -41 84 -104            u=1 imp:n=1 $ plate 15 161      3 -2.7        12 -10 64 -124 #160      u=1 imp:n=1 c 162      2 -1.0        12 -10 124 -19          u=1 imp:n=1 170      2 -1.0        -13:11:-50:124          u=1 imp:n=1 $ water between fuel and enclosure c
c      Universe 19: MIT with FHE c
201    0          30 38 39 fill=1              u=19 imp:n=1 202    5 -0.737    -33 39 -32 30                      u=19 imp:n=1 $ right neo 203    5 -0.737    31 32 30                      u=19 imp:n=1 $ left neo 204    3 -2.7      (-30:-31:32:33) 34 35 37      u=19 imp:n=1 $
enclosure 205    2 -0.8      -34:-35:36:37                      u=19 imp:n=1 $ water outside FHE c
c      Universe 20: FHE in tube (center) c 210    2 -0.9    -200            fill=19            u=20 imp:n=1 $ inside pipe 211    4 -7.94    200 -201                          u=20 imp:n=1 $ pipe 212    6 -0.096    201 -203 250 -251 252 -253        u=20 imp:n=1 $
insulation 213    0          203 250 -251 252 -253              u=20 imp:n=1 $ pipe to tube 214    4 -7.94    -250:251:-252:253                  u=20 imp:n=1 $ tube to inf c
c      Universe 21: FHE in tube (down) c 220    2 -0.9    -200            fill=19(2)          u=21 imp:n=1 $ inside pipe 221    4 -7.94    200 -201                          u=21 imp:n=1 $ pipe 6-157
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                      Rev. 16, May 2021 222    6 -0.096    201 -203 250 -251 252 -253 u=21 imp:n=1 $
insulation 223    0          203 250 -251 252 -253      u=21 imp:n=1 $ pipe to tube 224    4 -7.94    -250:251:-252:253          u=21 imp:n=1 $ tube to inf c
c      Universe 22: FHE in tube (up) c 230    2 -0.9    -200          fill=19(3)  u=22 imp:n=1 $ inside pipe 231    4 -7.94    200 -201                  u=22 imp:n=1 $ pipe 232    6 -0.096    201 -203 250 -251 252 -253 u=22 imp:n=1 $
insulation 233    0          203 250 -251 252 -253      u=22 imp:n=1 $ pipe to tube 234    4 -7.94    -250:251:-252:253          u=22 imp:n=1 $ tube to inf c
c      Universe 23: FHE in tube (right) c 240    2 -0.9    -200          fill=19(4)  u=23 imp:n=1 $ inside pipe 241    4 -7.94    200 -201                  u=23 imp:n=1 $ pipe 242    6 -0.096    201 -203 250 -251 252 -253 u=23 imp:n=1 $
insulation 243    0          203 250 -251 252 -253      u=23 imp:n=1 $ pipe to tube 244    4 -7.94    -250:251:-252:253          u=23 imp:n=1 $ tube to inf c
c      Universe 24: FHE in tube (left) c 250    2 -0.9    -200          fill=19(5)  u=24 imp:n=1 $ inside pipe 251    4 -7.94    200 -201                  u=24 imp:n=1 $ pipe 252    6 -0.096    201 -203 250 -251 252 -253 u=24 imp:n=1 $
insulation 253    0          203 250 -251 252 -253      u=24 imp:n=1 $ pipe to tube 254    4 -7.94    -250:251:-252:253          u=24 imp:n=1 $ tube to inf c
c      Universe 25: FHE in tube (up/right) c 260    2 -0.9    -200          fill=19(6)  u=25 imp:n=1 $ inside pipe 261    4 -7.94    200 -201                  u=25 imp:n=1 $ pipe 262    6 -0.096    201 -203 250 -251 252 -253 u=25 imp:n=1 $
insulation 263    0          203 250 -251 252 -253      u=25 imp:n=1 $ pipe to tube 264    4 -7.94    -250:251:-252:253          u=25 imp:n=1 $ tube to inf c
c      Universe 26: FHE in tube (up/left) c 6-158
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                              Rev. 16, May 2021 270    2 -0.9    -200            fill=19(7)          u=26 imp:n=1 $ inside pipe 271    4 -7.94    200 -201                          u=26 imp:n=1 $ pipe 272    6 -0.096    201 -203 250 -251 252 -253        u=26 imp:n=1 $
insulation 273    0          203 250 -251 252 -253              u=26 imp:n=1 $ pipe to tube 274    4 -7.94    -250:251:-252:253                  u=26 imp:n=1 $ tube to inf c
c      Universe 27: FHE in tube (down/right) c 280    2 -0.9    -200            fill=19(8)          u=27 imp:n=1 $ inside pipe 281    4 -7.94    200 -201                          u=27 imp:n=1 $ pipe 282    6 -0.096    201 -203 250 -251 252 -253        u=27 imp:n=1 $
insulation 283    0          203 250 -251 252 -253              u=27 imp:n=1 $ pipe to tube 284    4 -7.94    -250:251:-252:253                  u=27 imp:n=1 $ tube to inf c
c      Universe 28: FHE in tube (down/left) c 290    2 -0.9    -200            fill=19(9)          u=28 imp:n=1 $ inside pipe 291    4 -7.94    200 -201                          u=28 imp:n=1 $ pipe 292    6 -0.096    201 -203 250 -251 252 -253        u=28 imp:n=1 $
insulation 293    0          203 250 -251 252 -253              u=28 imp:n=1 $ pipe to tube 294    4 -7.94    -250:251:-252:253                  u=28 imp:n=1 $ tube to inf c
c      Universe 3: Array of Packages c
300  0    -300 301 -302 303 imp:n=1 u=3 lat=1 fill=-2:2 -2:2 0:0 25 25 22 26 26 25 25 22 26 26 23 23 20 24 24 27 27 21 28 28 27 27 21 28 28 10      px  2.5451  $  Al side 11      px  3.0226  $  Al side 12      px -2.5451  $  Al side 13      px -3.0226  $  Al side 18 10  py -3.02768  $  Al bottom 19 10  py 3.02768  $  Al top 20 10  py -3.34518  $  neoprene 21 10  py  3.34518  $  neoprene c
24      pz -28.41625  $ bottom of fuel 25      pz 28.41625    $ top of fuel (22.375")
30 20  p -1.71429 -1  0 -7.3152 $ inner FHE 31 21  p 1.71429 -1  0 -7.3152 $ inner FHE 32 21  p -1.71429 -1  0  7.3152 $ inner FHE 6-159
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report              Rev. 16, May 2021 33 20  p 1.71429 -1 0 7.3152 $  inner FHE 34 20  p -1.71429 -1 0 -7.9697 $ outer FHE 35 21  p 1.71429 -1 0 -7.9697 $  outer FHE 36 21  p -1.71429 -1 0 7.9697 $  outer FHE 37 20  p 1.71429 -1 0 7.9697 $  outer FHE 38 21  p 1.71429 -1 0 -6.6859 $  left neo 39 20  p 1.71429 -1 0 6.6859 $  right neo c
40    px -2.3878 $ meat width (w/2*cos(30))
41    px 2.3878 $ meat width c
50 10  py -4.34848 51 10  py -3.73888 52 10  py -3.12928 53 10  py -2.51968 54 10  py -1.91008 55 10  py -1.30048 56 10  py -0.69088 57 10  py -0.08128 58 10  py 0.52832 59 10  py 1.13792 60 10  py 1.74752 61 10  py 2.35712 62 10  py 2.96672 63 10  py 3.57632 64 10  py 4.18592 c
70 10  py -4.30530 71 10  py -3.69570 72 10  py -3.08610 73 10  py -2.47650 74 10  py -1.86690 75 10  py -1.25730 76 10  py -0.64770 77 10  py -0.03810 78 10  py 0.57150 79 10  py 1.18110 80 10  py 1.79070 81 10  py 2.40030 82 10  py 3.00990 83 10  py 3.61950 84 10  py 4.22910 c
90 10  py -4.22910 91 10  py -3.61950 92 10  py -3.00990 93 10  py -2.40030 94 10  py -1.79070 95 10  py -1.18110 96 10  py -0.57150 97 10  py 0.03810 98 10  py 0.64770 99 10  py 1.25730 100 10  py 1.86690 101 10  py 2.47650 102 10  py 3.08610 103 10  py 3.69570 6-160
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                  Rev. 16, May 2021 104 10  py 4.30530 c
110 10  py -4.18592 111 10  py -3.57632 112 10  py -2.96672 113 10  py -2.35712 114 10  py -1.74752 115 10  py -1.13792 116 10  py -0.52832 117 10  py 0.08128 118 10  py 0.69088 119 10  py 1.30048 120 10  py 1.91008 121 10  py 2.51968 122 10  py 3.12928 123 10  py 3.73888 124 10  py 4.34848 c
199    cz 6.9012 $ Al 200    cz 7.3838 $ IR pipe 201    cz 7.6581 $ OR pipe 203    cz 10.1981 $ 1" insulation c
250    px  -9.6032 $ square tube 251    px    9.6032 252    py  -9.6032 253    py    9.6032 c
300    px  10.033 $ lattice surfaces/sq. tube 301    px -10.033 302    py  10.033 303    py -10.033 310    px -50.165 $ 5x5 bounds 311    px  50.165 312    py -50.165 313    py  50.165 320    px -80.645 $ outer bounds 321    px  80.645 322    py -80.645 323    py  80.645 324    pz -58.8963 325    pz  58.8963 m2    1001.62c 2            $ water 8016.62c 1 mt2    lwtr.60t m3    13027.62c 1            $ Al m4    6000.66c    -0.08      $ SS-304 14000.60c -1.0 15031.66c -0.045 24000.50c -19.0 25055.62c -2.0 26000.55c -68.375 28000.50c -9.5 m5    1001.62c -0.056920    $ neoprene (no Cl) 6000.66c -0.542646 c      17000.66c -0.400434 6-161
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                Rev. 16, May 2021 m6      13027.62c  -26.5      $ insulation material 14000.60c  -23.4 8016.62c  -50.2 m10    92234.69c 2.3613E-05 $ fuel 92235.69c 3.6835E-03 92236.69c 1.3657E-05 92238.69c 1.9539E-04 13027.62c 5.0481E-02 c          total 5.4398E-02 c
*tr2    0 -0.7798 0 30 60 90 120  30 90  $ down
*tr3    0 0.7798 0 30 60 90 120    30 90  $ up
*tr4    0.7798 0 0                        $ right
*tr5    -0.7798 0 0                        $ left
*tr6    0.5514    0.5514 0                        $ up/right
*tr7    -0.5514    0.5514 0 90 0    90 180 90 90  $ up/left
*tr8    0.5514  -0.5514 0 90 0    90 180 90 90  $ down/right
*tr9    -0.5514  -0.5514 0                        $ down/left
*tr10    0 0 0 30 120 90 60 30 90  $ rotate fuel surfaces 30 deg CCW
*tr20    -0.7798 0 0 30.2 59.8 90  120.2 30.2 90 j j j -1 $ rotate right FHE 30.2 deg CCW
*tr21    0.7798 0 0 30.2 59.8 90  120.2 30.2 90 j j j -1 $ rotate left FHE 30.2 deg CCW c
mode  n kcode 2500 1.0 50 250 sdef    x=d1 y=d2 z=d3 si1    -50 50 sp1    0 1 si2    -50 50 sp2    0 1 si3    -31 31 sp3    0 1 6-162
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                        Rev. 16, May 2021 6.11 Appendix C: Criticality Analysis for Small Quantity Payloads The ATR FFSC may be utilized to transport fuel with a small U-235 fissile loading ( 400 g U-235). This fuel may be enriched up to 94% U-235. The intent is to bound in a generic manner several classes of research and development fuel types, as the geometry and fissile loading of such fuels is subject to change. These fuel types include those described in Section 1.2.2.4, Small Quantity Payload. In addition, some standard fuel elements, such as RINSC, classify for transport as a small quantity payload, as well as individual plates used to fabricate MURR, MIT, and Cobra fuel. The following analysis demonstrates that the ATR FFSC with small quantity payload fuel complies with the requirements of 10 CFR 71.55 and 71.59. Based on a 3x4 array of 10 undamaged packages and a 2x2 array of four damaged packages, the Criticality Safety Index (CSI), per 10 CFR 71.59, is 25.0.
6.11.1 Description of Criticality Design 6.11.1.1 Design Features Important for Criticality No special design features are required to maintain criticality safety. No poisons are utilized in the package. The separation provided by the packaging (outer flat-to-flat dimension of 7.9-in),
along with the limit on the number of packages per shipment, is sufficient to maintain criticality safety.
6.11.1.2 Summary Table of Criticality Evaluation The upper subcritical limit (USL) for ensuring that the ATR FFSC (single package or package array) is acceptably subcritical, as determined in Section 6.11.8, Benchmark Evaluations, is:
USL = 0.9209 The package is considered to be acceptably subcritical if the computed ksafe (ks), which is defined as keffective (keff) plus twice the statistical uncertainty (), is less than or equal to the USL, or:
ks = keff + 2  USL The USL is determined on the basis of a benchmark analysis and incorporates the combined effects of code computational bias, the uncertainty in the bias based on both benchmark-model and computational uncertainties, and an administrative margin. The results of the benchmark analysis indicate that the USL is adequate to ensure subcriticality of the package.
The packaging design is shown to meet the requirements of 10 CFR 71.55(b). Moderation by water in the most reactive credible extent is utilized in both the normal conditions of transport (NCT) and hypothetical accident conditions of transport (HAC) analyses. In the single package NCT models, full-density water fills the accessible cavity, while in the single package HAC models, full-density water fills all cavities. In all single package models, 12-in of water reflection is utilized.
A 3x4x1 array of 10 packages (2 empty locations) is utilized for the NCT array, while a 2x2x1 array of 4 packages is utilized in the HAC array. In the HAC array cases, partial moderation is considered to maximize array interaction effects. In all array models, 12-in of water reflection is utilized external to the array.
6-163
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                            Rev. 16, May 2021 The maximum results of the criticality calculations are summarized in Table 6.11-1. The maximum calculated ks is 0.8943, which occurs for the optimally moderated NCT array case.
The NCT array is more reactive than the HAC array because the NCT array is larger, and moderation is allowed in both conditions. In this case, the fuel mixture is modeled with a height of 32.5 cm, and void is modeled between the insulation and outer tube.
6.11.1.3 Criticality Safety Index The criticality safety index is defined in 10 CFR 71.59 as 50/N, where 5N packages are used in the NCT array configuration, and 2N packages are used in the HAC array configuration. A 2x2 array (2N = 4, or N = 2) is utilized for the HAC array calculations, while a 3x4 array of 10 packages (5N = 10, or N = 2) is utilized for the NCT array calculations. Therefore, the criticality safety index is 50/N = 50/2 = 25.0. With a CSI = 25.0, a maximum of four packages is allowed per exclusive use shipment.
Table 6.11 Summary of Small Quantity Payloads Criticality Evaluation Normal Conditions of Transport (NCT)
Case                                ks Single Unit Maximum                        0.6478 Array Maximum                            0.8943 Hypothetical Accident Conditions (HAC)
Case                                ks Single Unit Maximum                        0.7244 Array Maximum                            0.8222 USL = 0.9209 6.11.2 Fissile Material Contents The fissile material content is up to 400 g U-235 enriched up to 94% as a general payload material. Because HEU is modeled in the analysis, the results also apply to medium enriched uranium (MEU) and low enriched uranium (LEU) fuels. The analysis also applies to any generic fuel with U-235 as the fissile isotope. The objective is to bound research and development fuels with designs that are subject to change. The full list of anticipated contents bounded by this analysis is summarized in Section 1.2.2.4, Small Quantity Payload.
In general, for enrichments greater than 5% U-235, a system is more reactive using a homogenized mixture rather than an explicit heterogeneous representation11. Therefore, to simplify the modeling approach, the fuel is modeled as a homogenized mixture of uranium and water. Note that the homogenized representation is simply a conservative representation, and it is not implied that the actual fuel would behave in this manner. The fuel, even in accident conditions, would remain largely intact.
11 JJ Duderstadt and LJ Hamilton, Nuclear Reactor Analysis, p. 405, John Wiley & Sons, Inc., 1976.
6-164
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 This fuel mixture is assumed to conform to the cylindrical geometry constraint of the inner circular tube of the ATR FFSC. The fuel element structural materials (i.e., aluminum, silicon, etc.) are conservatively ignored, as well as the fuel handling enclosure (FHE) that supports the fuel element (either the RINSC FHE for RINSC fuel, or the small payload FHE for the remaining fuels). Modeling the structural materials would increase parasitic neutron absorption, as well as enlarge the size of the fissile volume to achieve the same hydrogen/U-235 ratio, and both effects would decrease the reactivity. A polyethylene limit of 100 g is justified in the analysis.
The contents may contain burnable absorbers, such as gadolinium, samarium, or boron. All burnable absorbers are conservatively neglected in the analysis.
The isotopic distribution of HEU fuel used in the analysis is listed in Table 6.11-2. The U-235 enrichment is conservatively modeled at 94%, which bounds the approximately 20% enrichment of LEU fuel, and 40-80% enrichment of MEU fuel. The remaining uranium isotopic values are representative and are consistent with the values used in the ATR criticality analysis (see Section 6.2, Fissile Material Contents). The fuel is modeled as homogenized mixture of uranium and water. Optimum reactivity is achieved by varying the height of the fissile mixture. A useful index of moderation for homogeneous systems is the hydrogen to U-235 ratio, abbreviated as H/U-235. This parameter is adjusted by varying the height of the fissile mixture. Increasing the height of the fissile mixture increases H/U-235.
The number densities of the homogenized mixture are computed in the following manner. A U-235 mass of 400 g is modeled, which bounds the masses of the small quantity payload items.
The weight percent of U-235 is 94.0%. Therefore, the total mass of uranium MU for 400 g U-235 is 400/0.94 = 425.5 g U. The theoretical density of uranium is 19.0 g/cm3, so the solid-volume VU of 425.5 g U is 425.5/19.0 = 22.4 cm3. The homogenized volume V is R2H, where R is the inner radius of the ATR FFSC circular tube (7.3838 cm) and H is the height of the fissile mixture. The gram density of uranium in the mixture is then MU/V, and if water of density 1.0 g/cm3 fills the remaining volume, the water density in the mixture is (V- VU)/V. The number densities of uranium and water may then be computed from the mixture densities. An example set of fuel mixture number densities for a height of 40 cm is provided in Table 6.11-3.
The ATR FFSC may contain hydrogenous materials. Fuel elements may be transported in a polyethylene (CH2) bag with a mass of approximately 3 oz, or 85 g. Neoprene (C4H5Cl) is used as a padding material in the fuel holders, and cellulosic material (C6H10O5) (e.g., kraft paper, cardboard) may be used as a cushioning material. The total mass of neoprene and cellulosic material is limited to a sum of 4000 g. Fiberglass reinforced tape may also be used to secure bundles of loose plates, and the mass of tape is conservatively treated as polyethylene.
Homogenized mixtures are developed that include either polyethylene, neoprene, cellulosic material, or structural material (such as aluminum) using the same method described above. For these computations, the density of polyethylene is 0.92 g/cm3, the density of neoprene is 1.23 g/cm3, the density of cellulosic material is 0.44 g/cm3, and the density of aluminum is 2.6989 g/cm3. As an example, fuel mixture number densities are provided in Table 6.11-3 for a height of 40 cm. Mixture number densities for other heights may be computed using the methodology described above.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                        Rev. 16, May 2021 Table 6.11 Uranium Isotopics Modeled HEU Isotope  Isotopics (Wt. %)
U-234          0.60 U-235          94.0 U-236          0.35 U-238          5.05 6-166
 
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                                    Rev. 16, May 2021 Table 6.11 Example Fissile Mixture Number Densities for Height = 40 cm Fuel/Water Number  Fuel/Water Number  Fuel/Water Number  Fuel/Water Number Fuel/Water    Densities (atom/b- Densities (atom/b- Densities (atom/b-  Densities (atom/b-Number Densities  cm) with 100 g of cm) with 1500 g of cm) with 1500 g of  cm) with 1500 g of Isotope    (atom/b-cm)      Polyethylene          Neoprene      Cellulosic Material    Aluminum U-234      9.5888E-07        9.5888E-07        9.5888E-07        9.5888E-07          9.5888E-07 U-235      1.4958E-04        1.4958E-04        1.4958E-04        1.4958E-04          1.4958E-04 U-236      5.5459E-07        5.5459E-07        5.5459E-07        5.5459E-07          5.5459E-07 U-238      7.9346E-06        7.9346E-06        7.9346E-06        7.9346E-06          7.9346E-06 H        6.6636E-02        6.6828E-02        6.2182E-02        4.1461E-02          6.1213E-02 O        3.3318E-02        3.2788E-02        2.7368E-02        2.0731E-02          3.0606E-02 C            -            6.2664E-04        5.9567E-03        4.8789E-03              -
Cl            -                  -            1.4892E-03              -                  -
Al            -                  -                  -                  -              4.8865E-03 Total    1.0011E-01        1.0040E-01        9.7155E-02        6.7230E-02          9.6864E-02 6-168
 
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 6.11.3 General Considerations 6.11.3.1 Model Configuration The packaging is modeled essentially the same as described in Section 6.3.1, Model Configuration. Refer to that section for details of the packaging model. The package length is modeled as 48-in long to be consistent with the original criticality models using ATR fuel (which has an active length of 48-in), although this length is somewhat arbitrary and is conservatively shorter than the actual inner cavity length of 67.88-in. The package is reflected with 12-in of full-density water.
In the NCT single package models, the inner tube, insulation, and outer tube are modeled explicitly, as shown in Figure 6.11-1 and Figure 6.11-2. Although negligible water ingress is expected during NCT, the inner cavity of the package is assumed to be flooded with water because the package lid does not contain a seal. However, the region between the insulation and the outer tube will remain dry because water cannot enter this region. The fuel is transported in a Fuel Handling Enclosure (FHE), which is conservatively ignored because the fuel is homogenized with water. Modeling the FHE would decrease the reactivity significantly if it is assumed that the fuel is homogenized within the constraint of the FHE. If it is assumed that the homogenized mixture could flow out of the FHE, modeling the FHE would still be less reactive than ignoring it because it would displace fissile material and increase the size of the fissile cylinder.
Although the FHE is not modeled, hydrogenous neoprene cushioning material along the sides of the enclosure is included in the fissile mixture in the NCT array models to demonstrate the poisoning effect of neoprene. The combined mass of neoprene and cellulosic material is limited to 4000 g.
The fuel elements may be transported in a polyethylene bag with an approximate mass of 3 oz, or 85 g. A polyethylene mass of 100 g is conservatively homogenized with the fuel/water mixture when indicated. The mass of fiberglass reinforced tape, which may be used to bind loose plates, shall be included in the polyethylene mass.
The HAC single package model is essentially the same as the NCT single package model.
Damage in the drop tests was shown to be negligible and concentrated at the ends of the package
[See Section 2.12.1, Certification Tests on CTU-1]. As the ends of the package are not modeled, this end damage does not affect the modeling. The various side drops resulted in only minor localized damage to the outer tube, and no observable bulk deformation of the package.
Therefore, the minor damage observed will not impact the reactivity. The insulation is replaced with full-density water, and the region between the insulation and outer tube is also filled with full-density water (see Figure 6.11-3). The treatment of the FHE is the same as the NCT single package model.
In the NCT array models, a 3x4x1 array is utilized, although two array positions are empty, for a total of 10 packages. The geometry of a package in the NCT array is the same as the NCT single package models. In the HAC array models, a 2x2x1 array is utilized. The HAC array models are essentially the same as the NCT array models, except additional cases are developed to determine the reactivity effect of allowing variable density water in the region between the inner and outer tubes. Cases are also developed with and without the insulation. The FHE is 6-170
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                Rev. 16, May 2021 conservatively ignored for the reasons stated in the previous paragraphs. Because the NCT and HAC models are very similar and the NCT models utilize a larger array, the NCT array models are more reactive than the HAC array models.
The detailed moderation assumptions for these cases are discussed more fully in Section 6.11.5, Evaluation of Package Arrays under Normal Conditions of Transport, and Section 6.11.6, Package Arrays under Hypothetical Accident Conditions.
6.11.3.2 Material Properties An example fissile material composition is provided in Table 6.11-3. The material properties of the packaging materials are provided in Section 6.3.2, Material Properties.
6.11.3.3 Computer Codes and Cross-Section Libraries The computer codes and cross-section libraries utilized are provided in Section 6.3.3, Computer Codes and Cross-Section Libraries.
6.11.3.4 Demonstration of Maximum Reactivity A number of conservative assumptions are utilized to obtain the maximum reactivity:
The fuel is modeled as a homogeneous mixture of uranium and water, which is a significantly more reactive configuration than modeling the fuel explicitly. Fuel element structural materials are ignored in the most reactive case.
400 g of U-235 is modeled, which bounds the U-235 loading of the proposed contents.
The U-235 enrichment is modeled as 94%, which bounds the enrichment of the proposed contents.
The fissile mixture is assumed to fill the inner tube of the ATR FFSC, and moderation is varied by running cases with different fissile mixture heights. No credit is taken for fuel handling enclosures that would maintain the fuel in a more favorable geometry. Note that the homogenized representation is simply a conservative representation, and it is not implied that the actual fuel would behave in this manner. The fuel, even in accident conditions, would remain largely intact.
In the NCT cases, water fills only the inner tube, because water would not enter the region between the inner circular tube and outer square tube. In the HAC cases, water is allowed in the region between the inner circular tube and outer square tube. Also, insulation may be replaced with water in the HAC cases. All single package cases are reflected with 12-in of water.
For the NCT array, 10 packages are modeled in a 3x4x1 array (with 2 empty locations), while in the HAC array, a smaller 2x2x1 array is utilized. Because negligible damage was observed in the drop tests, the package dimensions are the same between the NCT and HAC models.
Dimensions of the packaging are selected to maximize reactivity, and 12-in of close-water reflection is utilized.
The NCT array analysis is rather straightforward, because the only variable is the height of the fissile mixture. In the HAC array analysis, variables include the height of the fissile mixture, the presence or absence of insulation, and the water density of the region between the circular and square tubes. These parameters are varied to find the most reactive HAC condition.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                    Rev. 16, May 2021 Because fuel elements may be transported in polyethylene bags, 100 g of polyethylene is included in the fissile mixture. Polyethylene has a small, but positive, effect on the reactivity.
The hydrogenous materials neoprene and cellulosic material are shown to have a negative effect on reactivity because they are less effective at moderating the fissile mixture than the water that is displaced. Therefore, it is conservative to ignore neoprene and cellulosic material in the models. It is also explicitly demonstrated that modeling inert structural materials, such as aluminum, has a negative effect on the reactivity.
The NCT array is more reactive than the HAC array, primarily because the NCT array is significantly larger, and both cases use a homogenized fuel assumption. The most reactive NCT array case (Case HC16) has a fissile mixture height of 32.5 cm and results in a ks = 0.89427, which is below the USL of 0.9209. The most reactive HAC array case (Case HD34) results in a ks = 0.82217.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                    Rev. 16, May 2021 12-in water reflector 1-in insulation Fissile Mixture 6.03-in 7.9-in Figure 6.11 NCT Single Package Model (planar view) 6-173
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                  Rev. 16, May 2021 48-in Fissile Height Note that the ends of the package are conservatively treated simply as a water reflector.
Figure 6.11 NCT Single Package Model (axial view) 6-174
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                            Rev. 16, May 2021 Insulation and void replaced with water.
Figure 6.11 HAC Single Package Model (planar view) 6-175
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 6.11.4 Single Package Evaluation 6.11.4.1 Single Package Configuration 6.11.4.1.1      NCT Single Package Configuration The geometry of the NCT single package configuration is discussed in Section 6.11.3.1, Model Configuration. The fissile material is homogenized with water for a variety of fissile mixture heights. The water above the fissile mixture is modeled at full-density to maximize reflection.
The package is reflected with 12-in of water.
It is demonstrated in Section 6.11.5 that including neoprene, cellulosic material, or structural materials in the fissile mixture reduces the reactivity, so these materials are conservatively neglected in the NCT single package configuration.
Results are provided in Table 6.11-4. Cases HA1 through HA10 are without polyethylene, and Cases HA11 through HA20 include 100 g of polyethylene. The cases with polyethylene are slightly more reactive, although the effect is small. Maximum reactivity is achieved for Case HA13, with a fissile mixture height of 25.0 cm. The reactivity of this case is low, with ks =
0.64775. This result is below the USL of 0.9209.
6.11.4.1.2      HAC Singe Package Configuration The geometry of the HAC single package configuration is discussed in Section 6.11.3.1, Model Configuration. The fissile material is homogenized with water for a variety of fissile mixture heights. The water above the fissile mixture is modeled at full-density to maximize reflection.
The insulation is replaced with full-density water, and full-density water is also modeled between the inner and outer tubes. The package is reflected with 12-in of water.
It is demonstrated in Section 6.11.5 that including neoprene, cellulosic material, or structural materials in the fissile mixture reduces the reactivity, so these materials are conservatively neglected in the HAC single package configuration.
Results are provided in Table 6.11-5. Cases HB1 through HB10 are without polyethylene, and Cases HB11 through HB20 include 100 g of polyethylene. The cases with polyethylene are slightly more reactive, although the effect is small. Maximum reactivity is achieved for Case HB15, with a fissile mixture height of 27.5 cm. The reactivity of this case is low, with ks =
0.72441. This result is below the USL of 0.9209.
6.11.4.2 Single Package Results Following are the tabulated results for the single package cases. The most reactive configurations are listed in boldface.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                      Rev. 16, May 2021 Table 6.11 NCT Single Package Results Fissile Mixture                                  ks Case ID        Filename      Height (cm)          keff              (k+2)
No Polyethylene HA1        NS_HEU_H15            15.0          0.61609 0.00130      0.61869 HA2        NS_HEU_H20            20.0          0.63614 0.00115      0.63844 HA3        NS_HEU_H25            25.0          0.64046 0.00116      0.64278 HA4        NS_HEU_H275          27.5          0.64251 0.00114      0.64479 HA5        NS_HEU_H30            30.0          0.64189 0.00116      0.64421 HA6        NS_HEU_H325          32.5          0.63773 0.00111      0.63995 HA7        NS_HEU_H35            35.0          0.62944 0.00106      0.63156 HA8        NS_HEU_H40            40.0          0.62060 0.00105      0.62270 HA9        NS_HEU_H45            45.0          0.60913 0.00110      0.61133 HA10      NS_HEU_H50            50.0          0.59328 0.00104      0.59536 With 100 g Polyethylene HA11      NS_HEUP_H15          15.0          0.62298 0.00128      0.62554 HA12      NS_HEUP_H20          20.0          0.64179 0.00112      0.64403 HA13      NS_HEUP_H25          25.0          0.64531 0.00122      0.64775 HA14      NS_HEUP_H275          27.5          0.64503 0.00114      0.64731 HA15      NS_HEUP_H30          30.0          0.64193 0.00113      0.64419 HA16      NS_HEUP_H325          32.5          0.63741 0.00116      0.63973 HA17      NS_HEUP_H35          35.0          0.63154 0.00113      0.63380 HA18      NS_HEUP_H40          40.0          0.62058 0.00108      0.62274 HA19      NS_HEUP_H45          45.0          0.60798 0.00109      0.61016 HA20      NS_HEUP_H50          50.0          0.59553 0.00101      0.59755 6-177
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                      Rev. 16, May 2021 Table 6.11 HAC Single Package Results Fissile Mixture                                  ks Case ID        Filename      Height (cm)          keff              (k+2)
No Polyethylene HB1      HS_HEU_H15            15.0          0.69170 0.00124      0.69418 HB2      HS_HEU_H20            20.0          0.71519 0.00122      0.71763 HB3      HS_HEU_H225            22.5          0.72038 0.00131      0.72300 HB4      HS_HEU_H25            25.0          0.72067 0.00126      0.72319 HB5      HS_HEU_H275            27.5          0.71817 0.00114      0.72045 HB6      HS_HEU_H30            30.0          0.71422 0.00120      0.71662 HB7      HS_HEU_H325            32.5          0.70809 0.00116      0.71041 HB8      HS_HEU_H35            35.0          0.70653 0.00121      0.70895 HB9      HS_HEU_H40            40.0          0.69450 0.00111      0.69672 HB10      HS_HEU_H45            45.0          0.67855 0.00120      0.68095 With 100 g Polyethylene HB11      HS_HEUP_H15            15.0          0.69905 0.00128      0.70161 HB12      HS_HEUP_H20            20.0          0.71848 0.00128      0.72104 HB13      HS_HEUP_H225          22.5          0.72122 0.00125      0.72372 HB14      HS_HEUP_H25            25.0          0.72136 0.00120      0.72376 HB15      HS_HEUP_H275          27.5          0.72189 0.00126      0.72441 HB16      HS_HEUP_H30            30.0          0.71679 0.00130      0.71939 HB17      HS_HEUP_H325          32.5          0.71212 0.00123      0.71458 HB18      HS_HEUP_H35            35.0          0.70759 0.00119      0.70997 HB19      HS_HEUP_H40            40.0          0.69424 0.00111      0.69646 HB20      HS_HEUP_H45            45.0          0.67857 0.00112      0.68081 6-178
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 6.11.5 Evaluation of Package Arrays under Normal Conditions of Transport 6.11.5.1 NCT Array Configuration The NCT array model is a 3x4x1 array with two empty locations, for a total of 10 packages. The array configuration utilized is the most reactive 10 package configuration, with 9 packages in a 3x3 configuration, and one package at the center of a side, see Figure 6.11-4. Axial stacking configurations, such as 2x3x2 with two empty locations, would lower the reactivity and are not investigated. The geometry of the individual packages is the same as the NCT single package model. The entire array is reflected with 12-in of full-density water. Moderation is varied by adjusting the height of the fissile mixture. The region above the fissile mixture is filled with full density water to maximize reflection.
The following series of cases are run:
Table 6.11-6: Moderator as pure water or water with 100 g polyethylene Table 6.11-7: Moderator as water with 100 g, 1500 g, or 4000 g neoprene Table 6.11-8: Moderator as water with 100 g, 1500 g, or 4000 g cellulosic material Table 6.11-9: Moderator as water with 100 g or 1500 g aluminum The results for pure water and water with 100 g polyethylene are provided in Table 6.11-6. For a pure water moderator, ks = 0.89381 at a fissile mixture height of 30 cm (Case HC5). When 100 g of polyethylene is added, ks = 0.89427 at a fissile mixture height of 32.5 cm (Case HC16).
While polyethylene is a superior moderator than water, the results with polyethylene are statistically identical to the results with pure water because the mass of added polyethylene is small.
While neoprene and cellulosic material both contain hydrogen, these materials are less effective moderators than pure water and the reactivity decreases when these materials are added to the fissile mixture, as shown in Table 6.11-7 and Table 6.11-8. The chlorine in the neoprene also acts as a poison. Therefore, it is conservative to neglect neoprene and cellulosic material in the models. A combined mass limit for neoprene plus cellulosic material of 4000 g is therefore justified.
When aluminum is added to the fissile mixture, the reactivity also decreases, as shown in Table 6.11-9. There is a sizable quantity of aluminum within the package cavity due to both the fuel cladding and FHE structural materials. Therefore, the modeling approach is inherently conservative because all metallic structural materials are neglected. Inert materials in the fuel meat, such as molybdenum or silicon, are also conservatively neglected.
The most reactive condition is Case HC16, which includes 100 g polyethylene and has a fissile height of 32.5 cm. For this case, ks = 0.89427, which is below the USL of 0.9209.
6.11.5.2 NCT Array Results The results for the NCT array cases are provided in the following tables. The most reactive configurations are listed in boldface.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                            Rev. 16, May 2021 Table 6.11 NCT Array Results with Polyethylene Fissile Mixture                              ks Case ID      Filename        Height (cm)          keff          (k+2)
Moderator: Water HC1    NA_HEU_H15                15.0        0.81375  0.00130  0.81635 HC2    NA_HEU_H20                20.0        0.86031  0.00130  0.86291 HC3    NA_HEU_H25                25.0        0.88140  0.00120  0.88380 HC4    NA_HEU_H275              27.5        0.88591  0.00129  0.88849 HC5    NA_HEU_H30                30.0        0.89141  0.00120  0.89381 HC6    NA_HEU_H325              32.5        0.89089  0.00123  0.89335 HC7    NA_HEU_H35                35.0        0.89028  0.00114  0.89256 HC8    NA_HEU_H40                40.0        0.88126  0.00116  0.88358 HC9    NA_HEU_H45                45.0        0.87387  0.00116  0.87619 HC10    NA_HEU_H50                50.0        0.85981  0.00104  0.86189 Moderator: Water with 100 g Polyethylene HC11    NA_HEUP_H15              15.0        0.81856  0.00130  0.82116 HC12    NA_HEUP_H20              20.0        0.86138  0.00129  0.86396 HC13    NA_HEUP_H25              25.0        0.88386  0.00119  0.88624 HC14    NA_HEUP_H275              27.5        0.88818  0.00113  0.89044 HC15    NA_HEUP_H30              30.0        0.88873  0.00122  0.89117 HC16    NA_HEUP_H325              32.5        0.89207  0.00110  0.89427 HC17    NA_HEUP_H35              35.0        0.88988  0.00113  0.89214 HC18    NA_HEUP_H40              40.0        0.88439  0.00110  0.88659 HC19    NA_HEUP_H45              45.0        0.87352  0.00106  0.87564 HC20    NA_HEUP_H50              50.0        0.86011  0.00110  0.86231 6-180
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                        Rev. 16, May 2021 Table 6.11 NCT Array Results with Neoprene Fissile Mixture                                      ks Case ID      Filename      Height (cm)            keff                (k+2)
Moderator: Water with 100 g Neoprene HC30    NA_N100_H15              15.0          0.78812    0.00045      0.78902 HC31    NA_N100_H20              20.0          0.83258    0.00043      0.83344 HC32    NA_N100_H25              25.0          0.85690    0.00044      0.85778 HC33    NA_N100_H30              30.0          0.86748    0.00042      0.86832 HC34    NA_N100_H35              35.0          0.86790    0.00040      0.86870 HC35    NA_N100_H40              40.0          0.86268    0.00039      0.86346 HC36    NA_N100_H45              45.0          0.85398    0.00040      0.85478 HC37    NA_N100_H50              50.0          0.84194    0.00039      0.84272 Moderator: Water with 1500 g Neoprene HC40    NA_HEUN_H30              30.0          0.62595    0.00096      0.62787 HC41    NA_HEUN_H35              35.0          0.63410    0.00092      0.63594 HC42    NA_HEUN_H40              40.0          0.63992    0.00099      0.64190 HC43    NA_HEUN_H45              45.0          0.64188    0.00091      0.64370 HC44    NA_HEUN_H50              50.0          0.63611    0.00086      0.63783 HC45    NA_HEUN_H55              55.0          0.63358    0.00079      0.63516 HC46    NA_HEUN_H60              60.0          0.62747    0.00087      0.62921 Moderator: Water with 4000 g Neoprene HC90    NA_N4000_H40            40.0          0.42787    0.00023      0.42833 HC91    NA_N4000_H45            45.0          0.43470    0.00023      0.43516 HC92    NA_N4000_H50            50.0          0.43803    0.00024      0.43851 HC93    NA_N4000_H55            55.0          0.44046    0.00023      0.44092 HC94    NA_N4000_H60            60.0          0.44040    0.00023      0.44086 HC95    NA_N4000_H65            65.0          0.43867    0.00022      0.43911 HC96    NA_N4000_H70            70.0          0.43710    0.00022      0.43754 6-181
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 Table 6.11 NCT Array Results with Cellulosic Material Fissile Mixture                                      ks Case ID      Filename          Height (cm)          keff                (k+2)
Moderator: Water with 100 g Cellulosic Material HC50    NA_K100_H15                15.0          0.78017        0.00044    0.78105 HC51    NA_K100_H20                20.0          0.83462        0.00047    0.83556 HC52    NA_K100_H25                25.0          0.86375        0.00043    0.86461 HC53    NA_K100_H30                30.0          0.87624        0.00045    0.87714 HC54    NA_K100_H35                35.0          0.87873        0.00041    0.87955 HC55    NA_K100_H40                40.0          0.87524        0.00042    0.87608 HC56    NA_K100_H45                45.0          0.86633        0.00041    0.86715 HC57    NA_K100_H50                50.0          0.85509        0.00038    0.85585 Moderator: Water with 1500 g Cellulosic Material HC60    NA_K1500_H40              40.0          0.69615        0.00040    0.69695 HC61    NA_K1500_H45              45.0          0.72143        0.00039    0.72221 HC62    NA_K1500_H50              50.0          0.73626        0.00039    0.73704 HC63    NA_K1500_H55              55.0          0.74237        0.00038    0.74313 HC64    NA_K1500_H60              60.0          0.74390        0.00037    0.74464 HC65    NA_K1500_H65              65.0          0.74255        0.00037    0.74329 HC66    NA_K1500_H70              70.0          0.73673        0.00037    0.73747 HC67    NA_K1500_H75              75.0          0.72883        0.00034    0.72951 HC68    NA_K1500_H80              80.0          0.72050        0.00033    0.72116 Moderator: Water with 4000 g Cellulosic Material HC100    NA_K4000_H85              85.0          0.57976        0.00033    0.58042 HC101    NA_K4000_H90              90.0          0.58604        0.00033    0.58670 HC102    NA_K4000_H95              95.0          0.58907        0.00033    0.58973 HC103    NA_K4000_H100            100.0          0.59108        0.00031    0.59170 HC104    NA_K4000_H105            105.0          0.58916        0.00031    0.58978 HC105    NA_K4000_H110            110.0          0.58677        0.00031    0.58739 HC106    NA_K4000_H115            115.0          0.58307        0.00032    0.58371 6-182
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                        Rev. 16, May 2021 Table 6.11 NCT Array Results with Aluminum Fissile Mixture                              ks Case ID      Filename          Height (cm)        keff          (k+2)
Moderator: Water with 100 g Aluminum HC70    NA_A100_H15              15.0          0.80807 0.00047  0.80901 HC71    NA_A100_H20              20.0          0.85445 0.00045  0.85535 HC72    NA_A100_H25              25.0          0.87889 0.00043  0.87975 HC73    NA_A100_H30              30.0          0.88745 0.00043  0.88831 HC74    NA_A100_H325              32.5          0.88880 0.00043  0.88966 HC75    NA_A100_H35              35.0          0.88808 0.00040  0.88888 HC76    NA_A100_H40              40.0          0.88135 0.00041  0.88217 HC77    NA_A100_H45              45.0          0.87218 0.00038  0.87294 HC78    NA_A100_H50              50.0          0.86069 0.00038  0.86145 Moderator: Water with 1500 g Aluminum HC80    NA_A1500_H20              20.0          0.78043 0.00045  0.78133 HC81    NA_A1500_H25              25.0          0.82043 0.00044  0.82131 HC82    NA_A1500_H30              30.0          0.84233 0.00044  0.84321 HC83    NA_A1500_H35              35.0          0.85139 0.00042  0.85223 HC84    NA_A1500_H40              40.0          0.85197 0.00040  0.85277 HC85    NA_A1500_H45              45.0          0.84700 0.00038  0.84776 HC86    NA_A1500_H50              50.0          0.83791 0.00039  0.83869 HC87    NA_A1500_H55              55.0          0.82724 0.00036  0.82796 HC88    NA_A1500_H60              60.0          0.81461 0.00038  0.81537 6-183
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                        Rev. 16, May 2021 Figure 6.11 NCT Array Geometry 6-184
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 6.11.6 Package Arrays under Hypothetical Accident Conditions 6.11.6.1 HAC Array Configuration The HAC array model is a 2x2x1 array of the HAC single package model, as shown in Figure 6.11-5. Results are provided in Table 6.11-10. Because it has been demonstrated in the NCT single package, HAC single package, and NCT array cases that adding 100 g of polyethylene to the fissile mixture slightly increases the reactivity, all HAC array cases include 100 g of polyethylene.
In Cases HD1 through HD10, the region between the inner circular tube and outer square tube is filled with full-density water. Therefore, the insulation is replaced with water. The fissile mixture height is varied to find the optimum moderation, and the region above the fissile mixture is filled with full-density water. Of these 10 cases, Case HD4 is the most reactive, with a fissile mixture height of 25.0 cm.
In Cases HD11 through HD15, the most reactive fissile mixture height of 25.0 cm is modeled.
The insulation is modeled explicitly, and a range of water densities are modeled between the insulation and outer square tube. These cases are less reactive than Case HD4, indicating that it is conservative to ignore the insulation in the HAC array models.
In Cases HD16 through HD25, Case HD4 is modified for a range of water densities between the inner circular tube and outer square tube. Case HD22 is the most reactive, with ks = 0.81981 and a water density between tubes of 0.6 g/cm3. This case is slightly more reactive than Case HD4, for which ks = 0.81502. However, the reactivity gain by using a reduced water density between the tubes is small.
The most reactive fissile mixture height may change based on the water density between the tubes. For this reason, a limited number of additional cases are run for fissile mixture heights of 22.5 cm, 27.5 cm, and 30.0 cm. In Cases HD26 through HD31, the fissile mixture height is 22.5 cm and the water density is varied between 0.3 and 0.8 g/cm3. Cases HD32 through HD37 are similar except the fissile mixture height is 27.5 cm, and in Cases HD38 through HD43 the fissile mixture height is 30.0 cm. The most reactive case is Case HD34, which is slightly more reactive than Case HD22.
Therefore, Case HD34 is the most reactive, with ks = 0.82217. This case has a fissile mixture height of 27.5 cm, the insulation has been replaced with water, and the water density between the inner circular tube and outer square tube is 0.5 g/cm3. This case is below the USL of 0.9209.
Note that the most reactive HAC array case is less reactive than the most reactive NCT array case (Case HC16) because the NCT array uses 10 packages, while the HAC array uses only 4 packages.
6.11.6.2 HAC Array Results Following are the tabulated results for the HAC array cases. The most reactive configurations are listed in boldface.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                  Rev. 16, May 2021 Table 6.11 HAC Array Results Water Fissile Density Mixture Between Case                        Height    Tubes                                    ks ID        Filename          (cm)    (g/cm3)    Insulation  keff          (k+2)
HD1  HA_HEUP_H15              15.0      1.0        No      0.77954 0.00115  0.78184 HD2  HA_HEUP_H20              20.0      1.0        No      0.80655 0.00123  0.80901 HD3  HA_HEUP_H225            22.5      1.0        No      0.80899 0.00130  0.81159 HD4  HA_HEUP_H25              25.0      1.0        No      0.81254 0.00124  0.81502 HD5  HA_HEUP_H275            27.5      1.0        No      0.81232 0.00124  0.81480 HD6  HA_HEUP_H30              30.0      1.0        No      0.80789 0.00116  0.81021 HD7  HA_HEUP_H325            32.5      1.0        No      0.80247 0.00114  0.80475 HD8  HA_HEUP_H35              35.0      1.0        No      0.79682 0.00119  0.79920 HD9  HA_HEUP_H40              40.0      1.0        No      0.78144 0.00114  0.78372 HD10  HA_HEUP_H45              45.0      1.0        No      0.76909 0.00110  0.77129 HD11  HA_HEUP_H25_IW000        25.0      0        Yes    0.79417 0.00131  0.79679 HD12  HA_HEUP_H25_IW025        25.0    0.25        Yes    0.79759 0.00128  0.80015 HD13  HA_HEUP_H25_IW050        25.0    0.50        Yes    0.80131 0.00121  0.80373 HD14  HA_HEUP_H25_IW075        25.0    0.75        Yes    0.80017 0.00121  0.80259 HD15  HA_HEUP_H25_IW100        25.0      1.0        Yes    0.80331 0.00118  0.80567 HD16  HA_HEUP_H25_W000        25.0      0        No      0.79115 0.00126  0.79367 HD17  HA_HEUP_H25_W010        25.0      0.1        No      0.79794 0.00117  0.80028 HD18  HA_HEUP_H25_W020        25.0      0.2        No      0.80884 0.00133  0.81150 HD19  HA_HEUP_H25_W030        25.0      0.3        No      0.81008 0.00123  0.81254 HD20  HA_HEUP_H25_W040        25.0      0.4        No      0.81535 0.00116  0.81767 HD21  HA_HEUP_H25_W050        25.0      0.5        No      0.81666 0.00129  0.81924 HD22  HA_HEUP_H25_W060        25.0      0.6        No      0.81733 0.00124  0.81981 HD23  HA_HEUP_H25_W070        25.0      0.7        No      0.81576 0.00130  0.81836 HD24  HA_HEUP_H25_W080        25.0      0.8        No      0.81435 0.00121  0.81677 HD25  HA_HEUP_H25_W090        25.0      0.9        No      0.81266 0.00130  0.81526 HD26  HA_HEUP_H225_W030        22.5      0.3        No      0.80656 0.00134  0.80924 HD27  HA_HEUP_H225_W040        22.5      0.4        No      0.80968 0.00135  0.81238 HD28  HA_HEUP_H225_W050        22.5      0.5        No      0.81297 0.00126  0.81549 HD29  HA_HEUP_H225_W060        22.5      0.6        No      0.81408 0.00113  0.81634 HD30  HA_HEUP_H225_W070        22.5      0.7        No      0.81343 0.00114  0.81571 HD31  HA_HEUP_H225_W080        22.5      0.8        No      0.81282 0.00123  0.81528 HD32  HA_HEUP_H275_W030        27.5      0.3        No      0.81386 0.00118  0.81622 HD33  HA_HEUP_H275_W040        27.5      0.4        No      0.81679 0.00123  0.81925 HD34  HA_HEUP_H275_W050        27.5      0.5        No      0.81993 0.00112  0.82217 HD35  HA_HEUP_H275_W060        27.5      0.6        No      0.81757 0.00123  0.82003 HD36  HA_HEUP_H275_W070        27.5      0.7        No      0.81559 0.00114  0.81787 HD37  HA_HEUP_H275_W080        27.5      0.8        No      0.81315 0.00125  0.81565 (continued) 6-186
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                  Rev. 16, May 2021 Table 6.11 HAC Array Results (concluded)
Water Fissile Density Mixture Between Case                        Height    Tubes                                  ks ID        Filename          (cm)    (g/cm3)  Insulation  keff          (k+2)
HD38  HA_HEUP_H30_W030        30.0    0.3        No      0.81016 0.00115  0.81246 HD39  HA_HEUP_H30_W040        30.0    0.4        No      0.81437 0.00121  0.81679 HD40  HA_HEUP_H30_W050        30.0    0.5        No      0.81585 0.00121  0.81827 HD41  HA_HEUP_H30_W060        30.0    0.6        No      0.81631 0.00113  0.81857 HD42  HA_HEUP_H30_W070        30.0    0.7        No      0.81257 0.00108  0.81473 HD43  HA_HEUP_H30_W080        30.0    0.8        No      0.81328 0.00113  0.81554 6-187
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                        Rev. 16, May 2021 Without insulation With insulation Figure 6.11 HAC Array Geometry 6-188
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                        Rev. 16, May 2021 6.11.7 Fissile Material Packages for Air Transport See Section 6.7, which applies to all contents.
6.11.8 Benchmark Evaluations The Monte Carlo computer program MCNP5 v1.3012 is utilized for this benchmark analysis.
MCNP has been used extensively in criticality evaluations for several decades and is considered a standard in the industry.
The uranium isotopes utilize preliminary ENDF/B-VII cross section data that are considered by Los Alamos National Laboratory to be more accurate than ENDF/B-VI cross sections. ENDF/B-V cross sections are utilized for chromium, nickel, and iron because natural composition ENDF/B-VI cross sections are not available for these elements. The remaining isotopes utilize ENDF/B-VI cross sections. All cross sections utilized are at room temperature. A listing of the cross section libraries used in the ATR FFSC analysis is provided in Table 6.3-4. These cross sections are consistent with the cross sections utilized in the benchmarks.
The ORNL USLSTATS code13 is used to establish a USL for the analysis. USLSTATS provides a simple means of evaluating and combining the statistical error of the calculation, code biases, and benchmark uncertainties. The USLSTATS calculation uses the combined uncertainties and data to provide a linear trend and an overall uncertainty. Computed multiplication factors, keff, for the package are deemed to be adequately subcritical if the computed value of ks is less than or equal to the USL as follows:
ks = keff + 2  USL The USL includes the combined effects of code bias, uncertainty in the benchmark experiments, uncertainty in the computational evaluation of the benchmark experiments, and an administrative margin. This methodology has accepted precedence in establishing criticality safety limits for transportation packages complying with 10 CFR 71.
6.11.8.1 Applicability of Benchmark Experiments The critical experiment benchmarks are selected from the International Handbook of Evaluated Criticality Safety Benchmark Experiments14 based upon their similarity to the ATR FFSC and contents. The important selection parameters are high enriched uranium solutions with a thermal spectrum and no strong absorbers such as boron. Ten benchmarks are available that meet this criteria. Because this is a small benchmark set, to supplement these benchmark cases, an additional 45 benchmarks are used for high enriched uranium solutions with boron or cadmium, as well as 42 low enriched (10%) solutions without poisons. The titles for all utilized experiments are listed in Table 6.11-11.
12 MCNP5, MCNP - A General Monte Carlo N-Particle Transport Code, Version 5; Volume II: Users Guide, LA-CP-03-0245, Los Alamos National Laboratory, April, 2003.
13 USLSTATS, USLSTATS: A Utility To Calculate Upper Subcritical Limits For Criticality Safety Applications, Version 1.4.2, Oak Ridge National Laboratory, April 23, 2003.
14 OECD Nuclear Energy Agency, International Handbook of Evaluated Criticality Safety Benchmark Experiments, NEA/NSC/DOC(95)03, September, 2006.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 6.11.8.2 Bias Determination The USL is calculated by application of the USLSTATS computer program. USLSTATS receives as input the keff as calculated by MCNP, the total 1- uncertainty (combined benchmark and MCNP uncertainties), and a trending parameter. Two trending parameters have been selected: (1) Energy of the Average neutron Lethargy causing Fission (EALF), and (2) the ratio of the hydrogen to U-235 number density (H/U-235).
The uncertainty value, total, assigned to each case is a combination of the benchmark uncertainty for each experiment, bench, and the Monte Carlo uncertainty associated with the particular computational evaluation of the case, MCNP, or:
total = (bench2 + MCNP2)1/2 These values are input into the USLSTATS program in addition to the following parameters, which are the values recommended by the USLSTATS users manual:
P, proportion of population falling above lower tolerance level = 0.995 (note that this parameter is required input but is not utilized in the calculation of USL Method 1) 1-, confidence on fit = 0.95
    , confidence on proportion P = 0.95 (note that this parameter is required input but is not utilized in the calculation of USL Method 1) km, administrative margin used to ensure subcriticality = 0.05.
These data are followed by triplets of trending parameter value, computed keff, and uncertainty for each case. A confidence band analysis is performed on the data for each trending parameter using USL Method 1. The USL generated for each of the trending parameters utilized is provided in Table 6.11-12. All benchmark data used as input to USLSTATS are reported in Table 6.11-13.
Energy of the Average neutron Lethargy causing Fission (EALF)
The EALF is used as the first trending parameter for the benchmark cases. The EALF comparison provides a means to observe neutron spectral dependencies or trends. USLSTATS is run for all experiments, as well as the subset of experiments that do not contain poisons. The data for the subset of experiments without poisons are plotted in Figure 6.11-6, while the data for all experiments is plotted in Figure 6.11-7. Over the range of applicability, the minimum USL is 0.9344 for the subset of benchmarks that do not contain poisons, and is 0.9309 when all benchmarks are considered. In both cases the USL is trending downward for increasing EALF.
Note that for the benchmarks that do not contain poison, the data tests not normal by a small margin (chi = 12.4231, upper bound = 9.49). This behavior is judged to be acceptable, both because the deviation from normal is not large, and the USL generated from this data is bounded by the USL with poison.
EALF for all ATR FFSC small quantity payload cases falls within the range of applicability.
The EALF is 4.98E-08 MeV for the most reactive case (Case HC16).
H/U-235 Atom Ratio The H/U-235 atom ratio is used as the second trending parameter for the benchmark cases. The data for the subset of experiments without poisons are plotted in Figure 6.11-8, while the data for 6-190
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                Rev. 16, May 2021 all experiments is plotted in Figure 6.11-9. Over the range of applicability, the minimum USL is 0.9401 for the subset of benchmarks that do not contain poisons, and is 0.9359 when all benchmarks are considered. The USL is relatively constant over the range of applicability when no poisons are considered, and is trending downward for decreasing H/U-235 when all benchmarks are considered. Note that for the benchmarks that do not contain poison, the data tests not normal by a small margin (chi = 12.4231, upper bound = 9.49). This behavior is judged to be acceptable, both because the deviation from normal is not large, and the USL generated from this data is bounded by the USL with poison.
The H/U-235 atom ratio for all ATR FFSC small quantity payload cases falls within the range of applicability. The H/U-235 atom ratio is 363 for the most reactive case (Case HC16).
Recommended USL For the H/U-235 trending parameter, the minimum USL is 0.9359, while for the EALF trending parameter, the USL is 0.9309. Therefore, a USL of 0.9309 could be justified. However, a benchmark analysis was also performed for high-enriched plate fuel for the original ATR FFSC criticality analysis (see Section 6.8, Benchmark Evaluations). In that section, a USL of 0.9209 is justified. Therefore, a USL of 0.9209 is conservatively selected as the USL for this analysis for consistency.
6-191
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                        Rev. 16, May 2021 Table 6.11 Benchmark Experiments Utilized Series                                            Title HEU-SOL-THERM-001        Minimally Reflected Cylinders of Highly Enriched Solutions of Uranyl Nitrate HEU-SOL-THERM-027        Uranium (89% 235U) Nitrate Solution with Central Boron Carbide or Cadmium Absorber Rod HEU-SOL-THERM-028        Uranium (89% 235U) Nitrate Solutions with Central Boron Carbide Absorber Rod HEU-SOL-THERM-029        Uranium (89% 235U) Nitrate Solution with Cluster of Seven Boron Carbide Absorber Rods HEU-SOL-THERM-030        Uranium (89% 235U) Nitrate Solution with Cluster of Several Boron Carbide Absorber Rods HEU-SOL-THERM-036        Square-Pitched Lattices of Boron Carbide Absorber Rods In Uranium (89% 235U) Nitrate Solutions LEU-SOL-THERM-003        Full and Truncated Bare Spheres of 10% Enriched Uranyl Nitrate Water Solutions LEU-SOL-THERM-004        Stacy: Water-Reflected 10%-Enriched Uranyl Nitrate Solution in a 60-cm-Diameter Cylindrical Tank LEU-SOL-THERM-007        Stacy: Unreflected 10%-Enriched Uranyl Nitrate Solution in a 60-cm-Diameter Cylindrical Tank LEU-SOL-THERM-016        Stacy: 28-cm-Thick Slabs of 10%-Enriched Uranyl Nitrate Solutions, Water-Reflected LEU-SOL-THERM-017        Stacy: 28-cm-Thick Slabs of 10%-Enriched Uranyl Nitrate Solutions, Unreflected LEU-SOL-THERM-020        Stacy: 80-cm-Diameter Cylindrical Tank of 10%-Enriched Uranyl Nitrate Solutions, Water-Reflected LEU-SOL-THERM-021        Stacy: 80-cm-Diameter Cylindrical Tank of 10%-Enriched Uranyl Nitrate Solutions, Unreflected 6-192
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                Rev. 16, May 2021 Table 6.11 USL Results Minimum USL Trending      Experiment      Over Range of        Range of Parameter (X)        Set        Applicability    Applicability No poison (1-10, EALF (MeV)                            0.9344    3.43E-08  x  2.95E-07 56-97)
EALF (MeV)            All            0.9309    3.43E-08  x  2.95E-07 No poison (1-10, H/U-235                              0.9401        68.2  x  1437.5 56-97)
H/U-235                All            0.9359        68.2  x  1437.5 6-193
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                  Rev. 16, May 2021 Table 6.11 Benchmark Experiment Data EALF No      Case          k    mcnp      bench  total  (MeV)    H/U-235 1  HST001_C01    0.99661 0.00100    0.0060 0.0061  8.17E-08    181.8 2  HST001_C02    0.99185 0.00096    0.0072 0.0073  2.76E-07    70.6 3  HST001_C03    0.99921 0.00090    0.0035 0.0036  8.00E-08    185.7 4  HST001_C04    0.99586 0.00094    0.0053 0.0054  2.93E-07    68.2 5  HST001_C05    0.99785 0.00079    0.0049 0.0050  4.28E-08    499.4 6  HST001_C06    1.00159 0.00081    0.0046 0.0047  4.45E-08    458.8 7  HST001_C07    0.99693 0.00092    0.0040 0.0041  7.70E-08    193.3 8  HST001_C08    0.99696 0.00094    0.0038 0.0039  8.18E-08    181.8 9  HST001_C09    0.99087 0.00101    0.0054 0.0055  2.95E-07    68.2 10  HST001_C10    0.99005 0.00086    0.0054 0.0055  4.61E-08    427.4 11  HST027_C01    0.99609 0.00093    0.0046 0.0047  7.42E-08    203.6 12  HST027_C02    0.99522 0.00090    0.0043 0.0044  7.49E-08    203.6 13  HST027_C03    0.99626 0.00089    0.0037 0.0038  7.52E-08    203.6 14  HST027_C04    0.99780 0.00093    0.0037 0.0038  7.53E-08    203.6 15  HST027_C05    0.99563 0.00086    0.0044 0.0045  7.58E-08    203.6 16  HST027_C06    0.99028 0.00095    0.0043 0.0044  7.50E-08    203.6 17  HST027_C07    0.99604 0.00094    0.0038 0.0039  7.50E-08    203.6 18  HST027_C08    0.99772 0.00091    0.0035 0.0036  7.48E-08    203.6 19  HST027_C09    0.99517 0.00090    0.0039 0.0040  7.49E-08    203.6 20  HST028_C01    0.99350 0.00080    0.0023 0.0024  4.72E-08    374.6 21  HST028_C02    0.99332 0.00078    0.0034 0.0035  4.77E-08    374.6 22  HST028_C03    0.99596 0.00080    0.0026 0.0027  4.71E-08    374.6 23  HST028_C04    0.99814 0.00078    0.0028 0.0029  4.76E-08    374.6 24  HST028_C05    0.99070 0.00077    0.0031 0.0032  4.74E-08    374.6 25  HST028_C06    0.99492 0.00080    0.0023 0.0024  4.77E-08    374.6 26  HST028_C07    0.99497 0.00082    0.0038 0.0039  4.77E-08    374.6 27  HST028_C08    0.99433 0.00083    0.0027 0.0028  4.81E-08    374.6 28  HST028_C09    0.99179 0.00088    0.0049 0.0050  1.45E-07    91.5 29  HST028_C10    0.99032 0.00086    0.0053 0.0054  1.46E-07    91.5 30  HST028_C11    0.99179 0.00090    0.0051 0.0052  1.47E-07    91.5 31  HST028_C12    0.99009 0.00083    0.0046 0.0047  1.49E-07    91.5 32  HST028_C13    0.99102 0.00089    0.0058 0.0059  1.49E-07    91.5 33  HST028_C14    0.99180 0.00086    0.0046 0.0047  1.51E-07    91.5 34  HST028_C15    1.00006 0.00092    0.0064 0.0065  1.50E-07    91.5 35  HST028_C16    0.99561 0.00084    0.0052 0.0053  1.52E-07    91.5 36  HST028_C17    0.99144 0.00087    0.0066 0.0067  1.53E-07    91.5 37  HST028_C18    0.99322 0.00085    0.0060 0.0061  1.54E-07    91.5 38  HST029_C01    0.99468 0.00088    0.0066 0.0067  1.58E-07    91.5 (continued) 6-194
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                  Rev. 16, May 2021 Table 6.11 Benchmark Experiment Data EALF No      Case          k    mcnp      bench  total  (MeV)    H/U-235 39  HST029_C02    0.99722 0.00085    0.0058 0.0059  1.58E-07    91.5 40  HST029_C03    0.99112 0.00090    0.0068 0.0069  1.59E-07    91.5 41  HST029_C04    0.99158 0.00087    0.0074 0.0075  1.67E-07    91.5 42  HST029_C05    0.99602 0.00085    0.0067 0.0068  1.69E-07    91.5 43  HST029_C06    0.99484 0.00092    0.0065 0.0066  1.69E-07    91.5 44  HST029_C07    0.99381 0.00089    0.0063 0.0064  1.68E-07    91.5 45  HST030_C01    0.99405 0.00078    0.0039 0.0040  4.78E-08    374.6 46  HST030_C02    0.99786 0.00079    0.0032 0.0033  4.85E-08    374.6 47  HST030_C03    0.99465 0.00075    0.0031 0.0032  4.88E-08    374.6 48  HST030_C04    0.99533 0.00092    0.0064 0.0065  1.58E-07    91.1 49  HST030_C05    0.99334 0.00085    0.0058 0.0059  1.60E-07    91.1 50  HST030_C06    0.99430 0.00084    0.0059 0.0060  1.61E-07    91.1 51  HST030_C07    0.99458 0.00082    0.0064 0.0065  1.65E-07    91.1 52  HST036_C01    0.99355 0.00086    0.0045 0.0046  5.58E-08    302.5 53  HST036_C02    0.99779 0.00084    0.0039 0.0040  5.79E-08    302.5 54  HST036_C03    0.99834 0.00084    0.0044 0.0045  6.05E-08    302.5 55  HST036_C04    0.99971 0.00078    0.0062 0.0062  6.31E-08    302.5 56  LST003_C01    0.99621 0.00040    0.0039 0.0039  4.10E-08    770.3 57  LST003_C02    0.99383 0.00038    0.0042 0.0042  3.91E-08    877.6 58  LST003_C03    0.99926 0.00038    0.0042 0.0042  3.89E-08    897.0 59  LST003_C04    0.99292 0.00036    0.0042 0.0042  3.87E-08    913.2 60  LST003_C05    0.99641 0.00032    0.0048 0.0048  3.59E-08    1173.4 61  LST003_C06    0.99695 0.00030    0.0049 0.0049  3.57E-08    1213.1 62  LST003_C07    0.99535 0.00030    0.0049 0.0049  3.55E-08    1239.8 63  LST003_C08    0.99894 0.00027    0.0052 0.0052  3.45E-08    1411.6 64  LST003_C09    0.99697 0.00025    0.0052 0.0052  3.43E-08    1437.5 65  LST004_C01    1.00136 0.00067    0.0008 0.0010  4.17E-08    719.0 66  LST004_C29    1.00057 0.00065    0.0009 0.0011  4.08E-08    771.3 67  LST004_C33    0.99847 0.00059    0.0009 0.0011  3.96E-08    842.2 68  LST004_C34    1.00148 0.00061    0.0010 0.0012  3.88E-08    895.8 69  LST004_C46    1.00196 0.00052    0.0010 0.0011  3.82E-08    941.7 70  LST004_C51    0.99877 0.00056    0.0011 0.0012  3.78E-08    982.5 71  LST004_C54    1.00160 0.00052    0.0011 0.0012  3.73E-08    1017.5 72  LST007_C01    0.99414 0.00045    0.0009 0.0010  4.25E-08    709.2 73  LST007_C02    0.99734 0.00044    0.0009 0.0010  4.11E-08    770.0 74  LST007_C03    0.99472 0.00041    0.0010 0.0011  3.99E-08    842.2 75  LST007_C04    0.99791 0.00038    0.0011 0.0012  3.91E-08    896.0 76  LST007_C05    0.99628 0.00038    0.0011 0.0012  3.85E-08    942.2 77  LST016_C105    1.00345 0.00047    0.0013 0.0014  5.14E-08    468.7 78  LST016_C113    1.00438 0.00049    0.0013 0.0014  4.89E-08    514.2 79  LST016_C125    1.00368 0.00045    0.0014 0.0015  4.51E-08    608.4 80  LST016_C129    1.00225 0.00041    0.0014 0.0015  4.39E-08    650.2 (continued) 6-195
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                  Rev. 16, May 2021 Table 6.11 Benchmark Experiment Data (concluded)
EALF No      Case          k    mcnp      bench  total  (MeV)    H/U-235 81  LST016_C131    1.00227 0.00044    0.0014 0.0015  4.26E-08    699.1 82  LST016_C140    1.00142 0.00041    0.0015 0.0016  4.17E-08    738.9 83  LST016_C196    1.00218 0.00041    0.0015 0.0016  4.11E-08    771.8 84  LST017_C104    1.00273 0.00050    0.0013 0.0014  5.15E-08    468.7 85  LST017_C122    1.00223 0.00049    0.0013 0.0014  4.94E-08    510.8 86  LST017_C123    1.00095 0.00045    0.0014 0.0015  4.52E-08    610.9 87  LST017_C126    1.00158 0.00044    0.0014 0.0015  4.39E-08    650.1 88  LST017_C130    1.00164 0.00046    0.0015 0.0016  4.27E-08    699.2 89  LST017_C147    1.00152 0.00042    0.0015 0.0016  4.20E-08    729.0 90  LST020_C01    0.99867 0.00038    0.0010 0.0011  3.78E-08    971.0 91  LST020_C02    0.99796 0.00034    0.0010 0.0011  3.69E-08    1053.9 92  LST020_C03    0.99807 0.00033    0.0012 0.0012  3.60E-08    1168.0 93  LST020_C04    0.99839 0.00031    0.0012 0.0012  3.54E-08    1239.3 94  LST021_C01    0.99672 0.00038    0.0009 0.0010  3.79E-08    971.0 95  LST021_C02    0.99767 0.00035    0.0010 0.0011  3.71E-08    1052.7 96  LST021_C03    0.99630 0.00034    0.0011 0.0012  3.61E-08    1168.0 97  LST021_C04    0.99786 0.00032    0.0012 0.0012  3.57E-08    1238.9 6-196
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                          Rev. 16, May 2021 1.006 1.004 1.002 1
y = -23068x + 0.9998 0.998                                                              R2 = 0.1628 K-eff 0.996 0.994 0.992 0.99 0.988 0.00E+00      5.00E-08    1.00E-07  1.50E-07  2.00E-07    2.50E-07  3.00E-07  3.50E-07 EALF (MeV)
Figure 6.11 Benchmark Data Trend for EALF (no poisons) 1.006 1.004 1.002                                                y = -30627x + 0.9991 R2 = 0.2571 1
0.998 K-eff 0.996 0.994 0.992 0.99 0.988 0.00E+00  5.00E-08  1.00E-07  1.50E-07  2.00E-07    2.50E-07  3.00E-07  3.50E-07 EALF (MeV)
Figure 6.11 Benchmark Data Trend for EALF (all) 6-197
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 1.006 1.004 1.002 y = 4E-07x + 0.9982 1                                                          R2 = 0.0019 0.998 K-eff 0.996 0.994 0.992 0.99 0.988 0.0  200.0  400.0  600.0    800.0    1000.0  1200.0    1400.0      1600.0 H/U-235 Figure 6.11 Benchmark Data Trend for H/U-235 (no poisons) 1.006 1.004 y = 4E-06x + 0.9946 1.002                                                    R2 = 0.1975 1
0.998 K-eff 0.996 0.994 0.992 0.99 0.988 0.0  200.0  400.0  600.0    800.0    1000.0  1200.0    1400.0      1600.0 H/U-235 Figure 6.11 Benchmark Data Trend for H/U-235 (all) 6-198
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                          Rev. 16, May 2021 6.11.9 Sample Input Files A sample input file (NA_HEUP_H325) is provided for the most reactive case (Case HC16).
ATR Package 999      0        -320:321:-322:323:-324:325                      imp:n=0 900      0        310 -311 312 -313 24 -25      fill=3          imp:n=1 901      2 -1.0 (311:-310:313:-312:-24:25) 320 -321 322 -323 324 -325 imp:n=1 c
c        Universe 20: Fuel mixture with pipe c
200      10 1.0043E-01 200                      u=20 imp:n=1 $ fuel mix 201      2 -1.0            26 -200                    u=20 imp:n=1 $ water above fuel 202      4 -7.94        200 -201                      u=20 imp:n=1 $ pipe 203      6 -0.096      201 -203 250 -251 252 -253 u=20 imp:n=1 $ insulation 204      0              203 250 -251 252 -253        u=20 imp:n=1 $ insulation to tube 205      4 -7.94        -250:251:-252:253            u=20 imp:n=1 $ tube to inf c
c        Universe 21: Water c
210      2 -1.0        -204                          u=21 imp:n=1 c
c        Universe 3: Array of Packages c
300  0    -300 301 -302 303 imp:n=1 u=3 lat=1 fill=-1:1 -1:2 0:0 20 20 20 20 20 20 20 20 20 21 20 21 24      pz  0              $ bottom of fuel 25      pz  121.92          $ top of cavity (48")
26      pz  32.5            $ top of fuel mix c
200      cz  7.3838    $ IR pipe 201      cz  7.6581    $ OR pipe 203      cz  10.1981  $ 1" insulation 204      pz  1000      $ dummy c
250      px  -9.6032 $ square tube 251      px    9.6032 252      py  -9.6032 253      py    9.6032 c
300      px  10.033 $ lattice surfaces/sq. tube 301      px  -10.033 302      py  10.033 303      py  -10.033 310      px  -30.099 $ 3x4 bounds 311      px  30.099 312      py  -30.099 313      py  50.165 320      px  -60.579 $ outer bounds 321      px  60.579 6-199
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                              Rev. 16, May 2021 322    py -60.579 323    py 80.645 324    pz -30.48 325    pz 152.4 m2    1001.62c 2            $ water 8016.62c 1 mt2    lwtr.60t m3    13027.62c 1            $ Al m4    6000.66c    -0.08      $ SS-304 14000.60c -1.0 15031.66c -0.045 24000.50c -19.0 25055.62c -2.0 26000.55c -68.375 28000.50c -9.5 m5    1001.62c -0.056920 $ neoprene 6000.66c -0.542646 c      17000.66c -0.400434 m6    13027.62c -26.5        $ insulation material 14000.60c -23.4 8016.62c    -50.2 m10  92234.69c 1.1802E-06 $ HEU fuel H=32.5 M235=400.0 100g Poly 92235.69c 1.8410E-04 92236.69c 6.8258E-07 92238.69c 9.7657E-06 1001.62c 6.6822E-02 6000.66c 7.7125E-04 8016.62c 3.2640E-02 c        Total 1.0043E-01 mt10  lwtr.60t c
mode  n kcode 2500 1.0 50 250 sdef  x=d1 y=d2 z=d3 si1    -30 30 sp1    0 1 si2    -30 50 sp2    0 1 si3    0 32.5 sp3    0 1 6-200
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                              Rev. 16, May 2021 6.12 Appendix D: Criticality Analysis for the U-Mo Demonstration Element This analysis and any associated results have been removed as the ATR U-Mo demonstration element is no longer a possible payload.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                        Rev. 16, May 2021 6.13 Appendix E: Criticality Analysis for the Cobra Fuel Element The ATR FFSC may be utilized to transport a Cobra fuel element. The Cobra fuel element is a plate-type fuel similar to the ATR fuel element analyzed in this chapter, although the fuel geometry is different. The following analyses demonstrate that the ATR FFSC with a Cobra fuel element complies with the requirements of 10 CFR &sect;71.55 and &sect;71.59. Based on a 5x5 array of damaged packages, the Criticality Safety Index (CSI), per 10 CFR &sect;71.59, is 4.0.
6.13.1 Description of Criticality Design 6.13.1.1 Design Features No special design features are required to maintain criticality safety. No poisons are utilized in the package. The separation provided by the packaging (outer flat-to-flat dimension of 7.9-in),
along with the limit on the number of packages per shipment, is sufficient to maintain criticality safety.
6.13.1.2 Summary Table of Criticality Evaluation The upper subcritical limit (USL) for ensuring that the ATR FFSC (single package or package array) is acceptably subcritical, as determined in Section 6.13.9, Benchmark Evaluations, is:
USL = 0.9209 The package is considered to be acceptably subcritical if the computed ksafe (ks), which is defined as keffective (keff) plus twice the statistical uncertainty (), is less than or equal to the USL, or:
ks = keff + 2  USL The USL is determined on the basis of a benchmark analysis and incorporates the combined effects of code computational bias, the uncertainty in the bias based on both benchmark-model and computational uncertainties, and an administrative margin. The results of the benchmark analysis indicate that the USL is adequate to ensure subcriticality of the package.
The packaging design is shown to meet the requirements of 10 CFR 71.55(b)(d)(e). Moderation by water in the most reactive credible extent is utilized in both the normal conditions of transport (NCT) and hypothetical accident conditions of transport (HAC) analyses. In the single package NCT models, full-density water fills the accessible cavity, while in the single package HAC models, full-density water fills all cavities. In all single package and array models, 12-in of water reflection is utilized.
A 9x9x1 array of 81 packages is utilized for the NCT array, while a 5x5x1 array of 25 packages is utilized for the HAC array. In the NCT and HAC array cases, partial moderation is considered to maximize array interaction effects.
The maximum results of the criticality calculations are summarized in Table 6.13-1. The maximum calculated ks is 0.7643, which occurs for the optimally moderated HAC array case. In the HAC cases, pitch expansion is postulated as a bounding damaged fuel element condition, and the Cobra fuel handing enclosure is conservatively assumed to fail. The reactivity of the worst case is significantly below the USL of 0.9209.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                Rev. 16, May 2021 Table 6.13 Summary of Criticality Evaluation, Cobra Element Normal Conditions of Transport (NCT)
Case                            ks 71.55(b)(d): Single Unit Maximum                  0.4032 71.55(d), 71.59: 9x9x1 Array Maximum              0.7389 Hypothetical Accident Conditions (HAC)
Case                            ks 71.55(e): Single Unit Maximum                    0.4996 71.55(e), 71.59: 5x5x1 Array Maximum              0.7643 USL = 0.9209 6.13.1.3 Criticality Safety Index The criticality safety index is defined in 10 CFR 71.59 as 50/N, where 5N packages are used in the NCT array configuration, and 2N packages are used in the HAC array configuration. A 5x5 array (2N = 25, or N = 12.5) is utilized for the HAC array calculations, while a 9x9 array (5N =
81, or N = 16.2) is utilized for the NCT array calculations. Therefore, the criticality safety index is computed with the smaller value of N, or 50/N = 50/12.5 = 4.0. With a CSI = 4.0, a maximum of 25 packages are allowed per exclusive use shipment, or a maximum of 12 packages per non-exclusive use shipment.
6.13.2 Fissile Material Contents The fissile material content is either a single high-enriched uranium (HEU) or low-enriched uranium (LEU) Cobra fuel element. The package may transport only a single fuel element per package. The HEU fuel element has a U-235 enrichment < 94% and the fuel meat is uranium metal mixed with aluminum. The LEU fuel element has a U-235 enrichment < 20% and the fuel meat is U3Si2 mixed with aluminum. The HEU and LEU Cobra fuel elements are geometrically identical, differing only in the fuel meat composition and enrichment.
A Cobra fuel element consists of six concentric rings of fuel plates. Each ring is comprised of three curved plates and are joined by aluminum spacers, see Figure 6.13-1. The plates fit into slots in the spacers. The fuel plate cladding is aluminum.
The as-modeled isotopic distributions of HEU and LEU fuel are listed in Table 6.13-2. The U-235 enrichment for HEU is set to a bounding value of 94%, and the U-235 enrichment for LEU is set to a bounding value of 20%.
Because the fuel element is comprised of concentric plates, the inner plates are smaller and hence have a lower U-235 loading than the outer plates. The U-235 loading in each plate is provided in Table 6.13-3. Based on the stated tolerances on U-235 per plate and three plates per ring, the total U-235 loading per fuel element is 402.3 +/- 10.8 g U-235 for the HEU fuel element and 427.8
+/- 10.8 g U-235 for the LEU fuel element. In the MCNP models, the U-235 loading per plate is maximized, resulting in a conservative U-235 loading of 402.3 + 10.8 = 413.1 g U-235 for the HEU fuel element and 427.8 + 10.8 = 438.6 g U-235 for the LEU fuel element. The actual U-235 tolerance on the overall fuel element is +/- 8.0 g rather than +/- 10.8 g. Therefore, the as-6-203
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 modeled U-235 loading per fuel element is conservatively 10.8 - 8.0 = 2.8 g larger than the maximum allowed values.
The fuel element may contain burnable absorbers, such as gadolinium, samarium, or boron. All burnable absorbers are conservatively neglected in this analysis.
Each plate has a nominal fuel meat thickness of 0.063 cm and a nominal cladding thickness of 0.035 cm, or an overall plate thickness of 0.063 + 2*0.035 = 0.133 cm. Tolerances on the cladding and meat thickness are not defined, but the overall tolerance on the plate thickness is 0.005 cm. The active fuel length of each plate ranges from a minimum of 74.95 cm to a maximum of 77.45 cm. The inner radius and maximum fuel meat width for each plate are provided in Table 6.13-4.
To properly match the target U-235 loading in each plate, the density of the fuel meat is explicitly computed for each plate. The first step is to compute the fuel meat volume of each plate, which is the thickness*width*length. An example fuel meat volume calculation is provided in Table 6.13-5 for nominal active fuel length, nominal fuel meat thickness, and maximum width.
For HEU fuel, the density and composition of the fuel meat may be computed for each plate using the following method, which is summarized in Table 6.13-6. The U-235 density (g/cm3) for each plate is computed based on the U-235 mass and fuel meat volume in each plate. The U-235 density varies slightly between plates to preserve the target U-235 loading. The uranium density (g/cm3) is computed as the U-235 density/enrichment, where the enrichment is 0.94. The U-238 density (g/cm3) is then the uranium density minus U-235 density.
The total mixture density includes both uranium and aluminum. It is shown in the original ATR criticality analysis (see Table 6.2-2) that for high-enriched plate-type fuels, the total mixture density is approximately related to the U-235 density as 0.8733*(U-235 density) + 2.5357 g/cm3.
This equation is applied to the average U-235 density in the Cobra fuel element, and then the ratio of the total density to the U-235 density is treated as a constant. Using the data in Table 6.13-6, the U-235 average density is 0.9391 g/cm3, the corresponding mixture density is 3.3558 g/cm3, and the ratio of these two values is 3.3558/0.9391 = 3.5734. The total mixture density of each plate is then estimated as 3.5734*(U-235 density), and the aluminum density (g/cm3) is then the total density minus uranium density.
While the component densities differ for each plate, the weight percent of each component in the fuel mixture is a constant. For the example described above, the composition is summarized in Table 6.13-7.
For LEU fuel, the methodology is similar, although the fuel meat is a mixture of approximately 80 wt.% U3Si2 and 20 wt.% aluminum. The computation of the fuel meat density of each plate for a nominal active fuel length, nominal fuel meat thickness, and maximum width is provided in Table 6.13-8. The corresponding fuel meat composition, which is the same for each plate, is provided in Table 6.13-9.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                Rev. 16, May 2021 Table 6.13 Modeled Uranium Isotopics Isotope        HEU (Wt. %)    LEU (Wt. %)
U-235              94.0          20.0 U-238              6.0          80.0 Table 6.13 U-235 Loading Plate      HEU, U-235 (g)  LEU, U-235 (g) 1 (inner)        11.8 +/- 0.5    12.4 +/- 0.5 2          16.0 +/- 0.5      17.0 +/- 0.5 3          20.3 +/- 0.6      21.4 +/- 0.6 4          24.4 +/- 0.6      26.2 +/- 0.6 5          28.9 +/- 0.7      30.4 +/- 0.7 6 (outer)        32.7 +/- 0.7    35.2 +/- 0.7 Total        134.1 +/- 3.6    142.6 +/- 3.6 Fuel Element 402.3 +/- 10.8    427.8 +/- 10.8 (Total x 3)
Table 6.13 Plate Dimensions Inner Radius      Max Meat Plate            (cm)      Width (cm) 1 (inner)          1.598          2.800 2              2.030          3.705 3              2.463          4.615 4              2.896          5.525 5              3.328          6.435 6 (outer)          3.761          7.340 6-205
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                              Rev. 16, May 2021 Table 6.13 Fuel Meat Volume Example Nominal Fuel Meat Thickness    Nominal Active Max Fuel Meat    Fuel Meat Plate          (cm)          Length (cm)    Width (cm)    Volume (cm3) 1 (inner)        0.063            76.2          2.800            13.442 2            0.063            76.2          3.705            17.786 3            0.063            76.2          4.615            22.155 4            0.063            76.2          5.525            26.523 5            0.063            76.2          6.435            30.892 6 (outer)        0.063            76.2          7.340            35.236 6-206
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                                                                        Rev. 16, May 2021 Table 6.13 HEU Fuel Meat Densities Example A              B            C = A/B              D            E = C/D        F=E-C              G          H=G-E Fuel Meat        U-235            Enrich-          Uranium          U-238            Total      Aluminum Volume          Density            ment            Density        Density        Density        Density Plate        U-235 (g)        (cm3)          (g/cm3)        (U-235/U)          (g/cm3)        (g/cm3)        (g/cm3)        (g/cm3) 1 (inner)        12.3          13.442          0.9151            0.94            0.9735        0.0584        3.2699          2.2964 2            16.5          17.786          0.9277            0.94            0.9869        0.0592        3.3150          2.3281 3            20.9          22.155          0.9434            0.94            1.0036        0.0602        3.3710          2.3674 4            25.0          26.523          0.9426            0.94            1.0027        0.0602        3.3681          2.3654 5            29.6          30.892          0.9582            0.94            1.0193        0.0612        3.4239          2.4046 6 (outer)        33.4          35.236          0.9479            0.94            1.0084        0.0605        3.3871          2.3788 Average            -              -            0.9391                -                -              -          3.3558            -
Computed based on the average U-235 density as 0.8733*0.9391 + 2.5357 = 3.3558 g/cm3.
Computed as G = C*R, where C is the U-235 density and R is the ratio of the total density and the U-235 average density: R = 3.3558/0.9391 = 3.5734.
Table 6.13 HEU Fuel Meat Composition Example Composition Isotope                (Wt.%)
U-235                27.9850 U-238                  1.7863 Al                70.2290 6-207
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                                        Rev. 16, May 2021 Table 6.13 LEU Fuel Meat Densities G=
A                  B          C = A/B      D    E = C/D F=E-C                H=G-E E/0.9269 Meat          U-235  Enrich-  Uranium  U-238    U3Si2      Silicon Volume          Density    ment    Density Density  Density      Density Plate          U-235 (g)          (cm3)        (g/cm3)  (U-235/U) (g/cm3) (g/cm3)  (g/cm3)      (g/cm3) 1 (inner)          12.9            13.442          0.9597    0.2    4.7985  3.8388  5.1769      0.3784 2              17.5            17.786          0.9839    0.2    4.9195  3.9356  5.3075      0.3879 3              22.0            22.155          0.9930    0.2    4.9651  3.9721  5.3566      0.3915 4              26.8            26.523          1.0104    0.2    5.0522  4.0417  5.4505      0.3984 5              31.1            30.892          1.0067    0.2    5.0337  4.0270  5.4306      0.3969 6 (outer)          35.9            35.236          1.0188    0.2    5.0942  4.0753  5.4959      0.4017 I              J = G/I      K=J-G U3Si2 Weight Fraction in          Total      Aluminum Total          Density        Density Plate          Density            (g/cm3)        (g/cm3) 1 (inner)          0.8              6.4711          1.2942 2              0.8              6.6343          1.3269 3              0.8              6.6957          1.3391 4              0.8              6.8132          1.3626 5              0.8              6.7883          1.3577 6 (outer)          0.8              6.8698          1.3740 0.9269 is the weight fraction of uranium in U3Si2.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                            Rev. 16, May 2021 Table 6.13 LEU Fuel Meat Composition Composition Isotope      (Wt.%)
U-235        14.831 U-238        59.322 Al          20.000 Si          5.8473 6-209
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                        Rev. 16, May 2021 Figure 6.13 Cobra Geometry 6-210
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 6.13.3 General Considerations 6.13.3.1 Model Configuration The model configuration is relatively simple. Most packaging details are conservatively ignored, particularly at the ends. Because the package is long and narrow, array configurations will stack only in the lateral directions (e.g., 5x5x1). Therefore, the end details, for both the package and the fuel element, are conservatively ignored external to the active fuel region, and these end regions are simply modeled as full-density water. The package length is modeled to match the active length of the fuel element (30 +/- 0.5-in), while the actual inner cavity length is 67.88-in as shown on the packaging general arrangement drawing 60501-10. The package is reflected with 12-in of full-density water.
Tolerances on the packaging are selected to result in the most reactive condition, as described in Section 6.3.1, Model Configuration. The standard ATR FFSC MCNP models are utilized with no change to the packaging descriptions.
The package consists of two primary structural components, a circular inner tube and a square outer tube. The modeled tube OD is 6.03-in, the modeled wall thickness is 0.108-in, and the modeled tube ID is 5.814-in. The outer tube is modeled with a wall thickness of 0.169-in and outer dimension of 7.9-in.
The fuel element is transported in the Cobra Fuel Handling Enclosure (CFHE). The CFHE supports the fuel element outside the active fuel region and radially centers the fuel element both within the CFHE and within the package cavity. Neoprene pads are attached to the CFHE at the interface between the fuel element and CFHE, although these pads are well outside the active fuel region and are not included in the models. Neoprene also contains chlorine, which acts a neutron poison, and thus it is also conservative to ignore neoprene because it has a negative effect on reactivity. The wall of the CFHE is aluminum with a thickness of 0.19 +/- 0.06-in. The nominal inner diameter of the CFHE is 4.1-in. Both maximum and minimum thicknesses for the CFHE are considered in the models.
The dimensions used in the MCNP models as discussed in the preceding paragraphs are summarized in Table 6.13-10.
In the NCT single package models, the CFHE, inner tube, insulation, and outer tube are modeled explicitly, as shown in Figure 6.13-2 and Figure 6.13-3. Although negligible water ingress is expected during NCT, the inner cavity of the package is assumed to be flooded with water because the package lid does not contain a seal. However, the region between the insulation and the outer tube will remain dry because water cannot enter this region.
In the NCT array models, a 9x9x1 array is utilized. The CFHE would survive NCT events with negligible damage and the fuel element remains centered within the CFHE and package cavity.
Because the fuel elements may be transported in a thin (< 0.01-in) plastic bag, this plastic bag is assumed to act as a boundary for partial moderation effects. The plastic bag is not modeled explicitly because it is too thin to have an appreciable effect on the reactivity. Therefore, it is postulated that the fuel element channels may fill with full-density water, while the region between the fuel element and CFHE fills with variable density water. Axial movement of the fuel elements is not considered because axial movement would increase the effective active height of the system and reduce the reactivity due to increased neutron leakage.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                Rev. 16, May 2021 The HAC single package model is similar to the NCT single package model. Damage in the drop tests was shown to be negligible and concentrated at the ends of the package (see Section 2.12.1, Certification Tests on CTU-1). As the ends of the package are not modeled, this end damage does not affect the modeling. The various side drops resulted in only minor localized damage to the outer tube, and no observable bulk deformation of the package. Therefore, the minor damage observed will not impact the reactivity. The insulation is replaced with full-density water, and the region between the insulation and outer tube is also filled with full-density water (see Figure 6.13-4).
During HAC, it is postulated that the CFHE may become damaged. To bound all possible CFHE damage scenarios, HAC models are developed both with and without the CFHE (see Figure 6.13-4). Furthermore, because the ATR FFSC has not been drop tested with a Cobra fuel element, it is postulated that minor damage could occur to the Cobra fuel element. The Cobra fuel element utilizes aluminum cladding plates swaged into spacers, similar in design and construction materials to the ATR fuel element, which showed little damage after drop testing (see Section 2.12.1.5.2, ATR Fuel Element Inspection). To conservatively bound the potential fuel damage in the HAC models, the fuel plate pitch is allowed to expand uniformly until constrained by the inner diameter of the package. This pitch expansion increases the moderation and the reactivity. In actuality, such a large uniform expansion of the fuel plate pitch is not credible, and in the worst case scenario would be localized at one end of the fuel element. The modeled damage is intended to bound a fuel element that exhibits minor damage but is otherwise intact.
In the HAC array models, a 5x5x1 array is utilized. As with the HAC single package cases, postulated damage to the CFHE and fuel element are explicitly addressed. Cases are also developed to determine the reactivity effect of allowing variable density water in the region between the inner and outer tubes.
The detailed moderation assumptions for these cases are discussed more fully in Section 6.13.6, Evaluation of Package Arrays under Normal Conditions of Transport, and Section 6.13.7, Package Arrays under Hypothetical Accident Conditions.
6.13.3.2 Material Properties The methodologies for computing the fuel meat densities and compositions for HEU and LEU fuel are provided in Section 6.13.2, Fissile Material Contents.
The material properties of the packaging materials are provided in Section 6.3.2, Material Properties. Pure aluminum with a density of 2.7 g/cm3 is used for the fuel element cladding and CFHE wall. The inner and outer tubes of the package are constructed from stainless steel 304.
Insulation is modeled as Durablanket S insulation material.
6.13.3.3 Computer Codes and Cross-Section Libraries MCNP5 v1.30 is used for the criticality analysis. All cross sections utilized are at room temperature (293.6 K). The uranium isotopes utilize preliminary ENDF/B-VII cross section data that are more accurate than ENDF/B-VI cross sections. ENDF/B-V cross sections are utilized for chromium, nickel, and iron because natural composition ENDF/B-VI cross sections are not available for these elements. The remaining isotopes utilize ENDF/B-VI cross sections. Titles of the cross sections utilized in the models have been extracted from the MCNP output or the 6-212
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 users manual and are provided in Table 6.13-11. The S(,) card LWTR.60T is used to simulate hydrogen bound to water.
All cases are run with 5000 neutrons per generation for 850 generations, skipping the first 50.
The 1- uncertainty is less than 0.001 for all cases.
6.13.3.4 Demonstration of Maximum Reactivity The U-235 mass is maximized within each fuel plate and exceeds the U-235 fuel element limit allowed by the fuel specification by 2.8 g. It is demonstrated that the HEU fuel element bounds the LEU fuel element, and bounding calculations are performed for the HEU fuel element. The fuel element model is optimized to produce the maximum reactivity by minimizing the active fuel length, cladding thickness, and fuel meat thickness, while maximizing the arc length of the fuel meat, see Section 6.13.4, Most Reactive Fuel Element Model. Aluminum spacers that join the fuel plates are conservatively omitted to increase moderation.
The NCT and HAC array cases are significantly more reactive than the corresponding single package cases. Therefore, the following discussion is focused on the array cases.
For NCT, the fuel remains intact and is centered within the CFHE, which is also centered within the package cavity. Water moderation is allowed for all NCT cases because the package lid does not contain a seal, although full moderation under NCT is not credible because the package is dry loaded. A 9x9x1 array of packages is modeled for NCT. Because the fuel elements may be transported in a thin plastic bag, it is assumed the plastic bag may act as a boundary for partial moderation effects. In the NCT array cases, the water density is independently varied within the fuel element, between the fuel element and CFHE, and between the CFHE and the inner diameter of the package. The most reactive condition is with a fully flooded fuel element and CFHE and void between the CFHE and the inner diameter of the package. Case CENA20 is the most reactive NCT array case, with ks = 0.73887.
For HAC, drop test data on the Cobra fuel element is not available, although the ATR fuel element is a similar plate-fuel design that showed little damage from a 30 foot drop (see Section 2.12.1.5.2, ATR Fuel Element Inspection). As a bounding fuel damage scenario, the fuel element plate pitch is allowed to increase under HAC. The CFHE is also assumed to be damaged and is omitted from the models when the fuel element is modeled as damaged. A 5x5x1 array of packages is modeled for HAC. Optimum moderation is achieved with a pitch expansion of 0.5 cm and the fuel element fully moderated. All fuel elements in the array are shifted to the center of the array. Reactivity is increased slightly if 0.8 g/cm3 water is modeled between the fuel element and the inner diameter of the package. It is also demonstrated that reactivity is maximized if insulation between the inner and outer tube is modeled explicitly. Case CEHA68 is the most reactive HAC array case, with ks = 0.76431.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                Rev. 16, May 2021 Table 6.13 Key Model Dimensions Drawing Dimension          As-Modeled Parameter                  (in)              Dimension (in)
ATR FFSC Cavity length                      67.88                30 +/- 0.5 Package Length                      73.9                30 +/- 0.5 Inner tube outer diameter            6.0                  6.03 Inner tube thickness                0.12                  0.108 Outer tube flat-to-flat            8.0                    7.9 Outer tube thickness              0.188                  0.169 Insulation thickness                1.0                    1.0 CFHE Canister body ID                    4.1                    4.1 Canister wall thickness        0.19 +/- 0.06            0.19 +/- 0.06 Table 6.13 Cross Section Libraries Utilized Isotope/Element            MCNP ID                    Cross Section Label H-1                  1001.62c    1-h-1 at 293.6K from endf-vi.8 njoy99.50 C                    6000.66c    6-c-0 at 293.6K from endf-vi.6 njoy99.50 O-16                  8016.62c    8-o-16 at 293.6K from endf-vi.8 njoy99.50 Al-27                13027.62c    13-al-27 at 293.6K from endf-vi.8 njoy99.50 Si                  14000.60c    14-si-nat at 293.6K from endf/b-vi P-31                  15031.66c    15-p-31 at 293.6K from endf-vi.6 njoy99.50 Cr                  24000.50c    24-cr-nat at 293.6K from endf/b-v njoy Mn-55                  25055.62c    25-mn-55 at 293.6K from endf/b-vi.8 njoy99.50 Fe                  26000.55c    26-fe-nat at 293.6K from endf/b-v njoy Ni                  28000.50c    28-ni-nat at 293.6K from endf/b-v njoy U-235                  92235.69c    92-u-235 at 293.6K from t16 u235la9d njoy99.50 U-238                  92238.69c    92-u-238 at 293.6K from t16 u238la8h njoy99.50 6-214
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                      Rev. 16, May 2021 12-in water reflector 1-in insulation void                                                            CFHE 6.03-in 7.9-in Inner tube thickness = 0.108-in Outer tube thickness = 0.169-in Figure 6.13 NCT Single Package Model (planar view) 6-215
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                      Rev. 16, May 2021 Active fuel length (30-in nominal)
Note that the ends of the package are conservatively treated simply as a water reflector.
Figure 6.13 NCT Single Package Model (axial view) 6-216
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                            Rev. 16, May 2021 Intact CFHE, Undamaged Fuel Damaged (absent) CFHE, Damaged Fuel Figure 6.13 HAC Single Package Model (planar view) 6-217
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 6.13.4 Most Reactive Fuel Element Model Prior to performing the criticality analysis, a parametric study is performed to determine the most reactive fuel element model. The most reactive fuel element model is then used in subsequent models. Fuel element parameters considered are:
HEU vs. LEU fuel Active fuel length Cladding thickness Fuel meat thickness Fuel meat arc length (or width as a flat plate)
Presence or absence of aluminum spacers within the fuel In all cases, the maximum U-235 loading per plate is utilized, including tolerances, as provided in Table 6.13-3. The as-modeled U-235 per plate is summarized in Table 6.13-12. The total U-235 modeled per fuel element conservatively exceeds the maximum allowed by the fuel specification by 2.8 g U-235 for both HEU and LEU fuel.
The cases considered are summarized in Table 6.13-13. Case CEP01 is the base case and features HEU fuel, nominal active fuel length, nominal cladding thickness, nominal fuel meat thickness, maximum arc length, and no aluminum spacers within the fuel. The thickness of the CFHE is modeled at the nominal value of 0.19-in in all models. In Cases CEP02 through CEP10, a single parameter is varied from the base case to determine the effect on the reactivity.
Case CEP11 is a combination of the worst-case parameters from Cases CEP01 through CEP10 and is the fuel element model used in subsequent analysis.
The considered variations are:
HEU fuel vs. LEU fuel is considered.
Active fuel length o Nominal = 76.2 cm o Maximum = 77.45 cm o Minimum = 74.95 cm Cladding thickness o Nominal = 0.035 cm o Maximum = 0.005 tolerance assumed, 0.035 + 0.005 = 0.04 cm o Minimum = 0.005 tolerance assumed, 0.035 - 0.005 = 0.03 cm Fuel meat thickness o Nominal = 0.063 cm o Maximum = 0.005 tolerance assumed, 0.063 + 0.005 = 0.068 cm o Minimum = 0.005 tolerance assumed, 0.063 - 0.005 = 0.058 cm Fuel meat arc length o Maximum = see Table 6.13-14 o Minimum = -5% tolerance on maximum values assumed, see Table 6.13-14 Aluminum spacers o The base model does not include aluminum spacers so that a water gap exists between the three fuel plates in each ring (see upper half of Figure 6.13-5) o A model is developed that includes the aluminum spacers between the three fuel plates in each ring (see lower half of Figure 6.13-5) 6-218
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 The base case (Case CEP01) is developed so that the inner radius of each plate matches the nominal values provided in Table 6.13-15. The centerline radius of each plate is the inner radius
+ cladding thickness + 1/2 fuel meat thickness and is also provided in Table 6.13-15. For simplicity, the centerline radius of each plate as defined in Table 6.13-15 is treated as a fixed quantity for all models in the parametric study. Therefore, adjustments to the cladding or fuel meat thicknesses do not affect the centerline radii. To properly preserve the volume of the fuel meat when a flat plate is bent into a curved plate, the arc length of the plate is related to the centerline radius based on the geometrical relationship S = r, where S is the fuel meat arc length and r is the centerline radius of the fuel meat. This allows computation of  for each plate, which is used to define the planes that terminate the edge of the plates.
Because the U-235 loading in each plate is fixed at the maximum value, any adjustments to the fuel meat thickness, fuel meat arc length, or active fuel length changes the composition and densities of the fuel meat in the plates. The methodology for computing the fuel meat densities and composition is provided in Section 6.13.2, Fissile Material Contents. The composition and densities for each of the cases is summarized in Table 6.13-16 and Table 6.13-17 for HEU and LEU fuel, respectively.
The aluminum spacers between the fuel plates are included in Case CEP09. The spacers are modeled with a thickness of 0.558 cm. For simplicity, the spacers are modeled with an ID and OD to match the overall fuel element boundary, 2.58 cm ID and 8.26 cm OD.
In this parametric study, packages are modeled in a 9x9x1 NCT array to exaggerate the reactivity effect one would observe from a single package. The interior of the package cavity and CFHE are filled with full-density water.
Results are provided in Table 6.13-18. The difference in ks between the base case and the perturbed case is provided in the far right column. This difference is expressed as ks*103.
Because the U-235 mass is fixed at the same value for each case, reducing the active fuel length has a small positive effect on reactivity due to decreased neutron leakage (compare Cases CEP03 and CEP01). Reducing the fuel meat arc length (Case CEP04) decreases the reactivity, consistent with the results for ATR fuel (see Section 6.4.1.2.1, Fuel Element Payload Parametric Evaluation). Decreasing the cladding thickness (compare Cases CEP05 and CEP01) has a small positive effect on reactivity because the width of the water channel between fuel plates increases, which increases moderation. Likewise, decreasing the fuel meat thickness has a positive effect on reactivity (compare Cases CEP08 and CEP01) because this reduces the overall thickness of the fuel plate, increases the width of the water channel, and increases moderation. Adding the aluminum spacers between the plates (Case CEP09) has a large negative effect on the reactivity due to decreased moderation, and it is conservative to omit the spacers from the model.
The LEU model (Case CEP10) is developed based on Case CEP01 fuel element geometry. The LEU case is less reactive than the base HEU case despite the larger U-235 loading in LEU fuel because LEU fuel has increased parasitic neutron absorption in U-238. Therefore, it is sufficient to perform the Cobra criticality analysis for the HEU fuel element, as it bounds the LEU fuel element.
Considering the results of Cases CEP01 through CEP10, the most reactive fuel element consists of HEU fuel meat, minimum active fuel length, minimum cladding thickness, minimum fuel meat thickness, maximum fuel meat arc length, and no fuel element spacers. Case CEP11 has 6-219
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 these parameters and results in the largest increase in reactivity over the base case, or 7.35 ks*103. The geometry and material description of Case CEP11 is used in all subsequent cases.
Table 6.13 U-235 Loading with Maximum Tolerance HEU, Max            LEU, Max Plate                U-235 (g)          U-235 (g) 1 (inner)                12.3              12.9 2                      16.5              17.5 3                      20.9              22.0 4                      25.0              26.8 5                      29.6              31.1 6 (outer)                33.4              35.9 Total                  137.7              146.2 Fuel Element 413.1              438.6 (Total x 3)
Fuel Specification 410.3              435.8 Maximum Table 6.13 Cases for Parameter Study Fuel HEU/        Active      Cladding    Fuel Meat    Fuel Meat    Element Case ID      LEU        length      Thickness    Thickness  Arc Length    Spacers CEP01        HEU          Nom            Nom          Nom          Max          No CEP02        HEU          Max          Nom          Nom          Max          No CEP03        HEU          Min          Nom          Nom          Max          No CEP04        HEU          Nom            Nom          Nom          Min          No CEP05        HEU          Nom            Max          Nom          Max          No CEP06        HEU          Nom            Min          Nom          Max          No CEP07        HEU          Nom            Nom          Max          Max          No CEP08        HEU          Nom            Nom          Min          Max          No CEP09        HEU          Nom            Nom          Nom          Max          Yes CEP10        LEU        Nom            Nom          Nom          Max          No Case CEP11 combines the worst-case parameters of the cases above CEP11        HEU          Min          Min          Min          Max          No Note: Changes from base case CEP01 highlighted in yellow.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                            Rev. 16, May 2021 Table 6.13 Fuel Meat Arc Length Plate          Maximum (cm)        Minimum (cm) 1 (inner)              2.800                2.660 2                  3.705                3.520 3                  4.615                4.384 4                  5.525                5.249 5                  6.435                6.113 6 (outer)              7.340                6.973 Table 6.13 Plate Radii Nominal Inner          Centerline Radius of Plate      Radius of Plate Plate                (cm)            (Fixed) (cm) 1 (inner)            1.5980              1.6645 2                2.0300              2.0965 3                2.4630              2.5295 4                2.8960              2.9625 5                3.3280              3.3945 6 (outer)            3.7610              3.8275 Table 6.13 HEU Fuel Meat Densities and Compositions CEP01 CEP05 CEP06 Case ID      CEP09      CEP02      CEP03        CEP04      CEP07    CEP08      CEP11 Plate                            Total Fuel Meat Density  (g/cm3) 1 (inner)    3.2699      3.2570    3.2832        3.3119    3.2111    3.3388      3.3532 2        3.3150      3.3019    3.3285        3.3576    3.2554    3.3848      3.3995 3        3.3710      3.3577    3.3847        3.4143    3.3104    3.4420      3.4569 4        3.3681      3.3549    3.3819        3.4115    3.3076    3.4391      3.4540 5        3.4239      3.4104    3.4379        3.4680    3.3624    3.4961      3.5112 6 (outer)    3.3871      3.3738    3.4009        3.4307    3.3263    3.4585      3.4735 Isotope                              Fuel Meat Composition (wt. %)
U-235      27.9850    27.6420    28.3360      29.0840    26.4020    29.7700    30.1360 U-238      1.7863      1.7644    1.8087        1.8564    1.6852    1.9002      1.9236 Al      70.2290    70.5930    69.8550      69.0600    71.9130    68.3300    67.9410 6-221
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                        Rev. 16, May 2021 Table 6.13 LEU Fuel Meat Densities and Composition Case ID                CEP10 Total Fuel Meat Density Plate                (g/cm3) 1 (inner)              6.4711 2                  6.6343 3                  6.6957 4                  6.8132 5                  6.7883 6 (outer)              6.8698 Fuel Meat Composition Isotope                (wt. %)
U-235                14.8310 U-238                59.3220 Al                20.0000 Si                  5.8473 Table 6.13 Parametric Study Results Case                                                              ks        k*103 from ID        Filename              keff                      (k+2)          Base CEP01  NA_G1                    0.67630          0.00040      0.67710          Base CEP02  NA_G2                    0.67399          0.00038      0.67475          -2.35 CEP03  NA_G3                    0.67652          0.00038      0.67728            0.18 CEP04  NA_G4                    0.67308          0.00039      0.67386          -3.24 CEP05  NA_G5                    0.67187          0.00036      0.67259          -4.51 CEP06  NA_G6                    0.67964          0.00038      0.68040            3.30 CEP07  NA_G7                    0.67359          0.00038      0.67435          -2.75 CEP08  NA_G8                    0.67724          0.00039      0.67802            0.92 CEP09  NA_G1_AL                0.65216          0.00038      0.65292          -24.18 CEP10  NA_G1_LEU                0.66761          0.00039      0.66839          -8.71 CEP11  NA_GMAX                  0.68369          0.00038      0.68445          7.35 6-222
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                            Rev. 16, May 2021 Case CEP1: No aluminum spacers Case CEP9: With aluminum spacers Figure 6.13 Fuel Element Models 6-223
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                Rev. 16, May 2021 6.13.5 Single Package Evaluation The most reactive fuel element is determined in Section 6.13.4, Most Reactive Fuel Element Model and consists of HEU fuel meat, minimum active fuel length, minimum cladding thickness, minimum fuel meat thickness, maximum fuel meat arc length, and no fuel element spacers. A fuel element with these parameters is used in the following analysis.
6.13.5.1 Single Package Configuration 6.13.5.1.1    NCT Single Package Configuration The geometry of the NCT single package configuration is discussed in Section 6.13.3.1, Model Configuration, and depicted in Figure 6.13-2. A single fuel element is modeled in the center of the CFHE. The package is reflected with 12-in of water.
Results are provided in Table 6.13-19. The first two cases feature a fully flooded package cavity and CFHE. Case CENS1 features a CFHE with maximum thickness (0.25-in), while Case CENS2 features a CFHE with minimum thickness (0.13-in). For a single package, reactivity is maximized for a minimum thickness CFHE because water reflection is enhanced.
In Cases CENS3 through CENS7, the water density within the fuel element is reduced. Full-density water is modeled between the fuel element and CFHE to maximize reflection. Full-density water is also modeled between the CFHE and package inner diameter to maximize reflection. Reactivity decreases as the water density within the fuel element is reduced.
Maximum reactivity is achieved for Case CENS2. This case is fully-flooded with a minimum thickness CFHE. The reactivity of this case is low, with ks = 0.40321. This case meets the requirements of 71.55(b)(d) and is below the USL of 0.9209.
6.13.5.1.2    HAC Single Package Configuration The geometry of the HAC single package configuration is discussed in Section 6.13.3.1 and is depicted in Figure 6.13-4. The package is reflected with 12-in of water.
Results for undamaged fuel are provided in Table 6.13-20. Only fully flooded cases are considered because it is demonstrated in the NCT single package cases that reactivity is reduced when the fuel element is moderated with lower density water. In Case CEHS1, the CFHE is modeled at the minimum thickness to maximize water reflection. Reactivity increases compared to the equivalent NCT single package case (Case CENS2) because insulation and void between the inner and outer tubes is replaced with water to increase reflection. In Case CEHS2, the CFHE is omitted from the model entirely to simulate damage to the CFHE. Reactivity increases when the CFHE is omitted from the model due to the increased water reflection.
Results for damaged fuel are provided in Table 6.13-21. The CFHE is conservatively omitted from all damaged fuel models to enhance water reflection and to allow more volume for pitch expansion. In the initial damaged fuel element model (Case CEHS10), the fuel element is compressed by shifting all fuel plates inward by 0.35 cm. In all subsequent damaged fuel element cases, the inner ring of fuel is fixed at the compressed location, while rings 2 through 6 are shifted outward from the base compressed case in 0.1 cm increments. Illustrations of damaged fuel element geometry for several of the damaged fuel cases are provided in Figure 6.13-6.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 The reactivity increases with increasing pitch, reaching a maximum ks = 0.49961 for a pitch expansion of 0.5 cm (Case CEH15). While the pitch expansion has a positive effect on reactivity, the reactivity remains relatively low. This case meets the requirements of 71.55(e) and is below the USL of 0.9209.
6.13.5.2 Single Package Results Following are the tabulated results for the single package cases. The most reactive configurations are listed in boldface.
Table 6.13 NCT Single Package Results Water Density within Fuel                                    ks Case ID            Filename            Element (g/cm3)      keff                  (k+2)
CENS1      NS_W100_MAXFHE                  1.0          0.39018      0.00034      0.39086 CENS2      NS_W100_MINFHE                  1.0          0.40249      0.00036      0.40321 CENS3      NS_W050_MINFHE                  0.5          0.28811      0.00030      0.28871 CENS4      NS_W060_MINFHE                  0.6          0.31051      0.00030      0.31111 CENS5      NS_W070_MINFHE                  0.7          0.33316      0.00033      0.33382 CENS6      NS_W080_MINFHE                  0.8          0.35609      0.00034      0.35677 CENS7      NS_W090_MINFHE                  0.9          0.37976      0.00034      0.38044 Table 6.13 HAC Single Package Results, Undamaged Fuel ks Case ID        Filename          keff                  (k+2)
CEHS1      HS_FHE              0.43296      0.00034      0.43364 CEHS2      HS_NOFHE            0.44386      0.00036      0.44458 Table 6.13 HAC Single Package Results, Damaged Fuel ks Case ID        Filename          Change in Pitch        keff                  (k+2)
CEHS10      HS_NOFHE_P0      Base Compressed          0.42535      0.00033      0.42601 CEHS11      HS_NOFHE_P1      0.1 cm from CEHS10      0.45449      0.00035      0.45519 CEHS12      HS_NOFHE_P2      0.2 cm from CEHS10      0.47424      0.00037      0.47498 CEHS13      HS_NOFHE_P3      0.3 cm from CEHS10      0.48837      0.00034      0.48905 CEHS14      HS_NOFHE_P4      0.4 cm from CEHS10      0.49530      0.00036      0.49602 CEHS15      HS_NOFHE_P5      0.5 cm from CEHS10      0.49889      0.00036      0.49961 CEHS16      HS_NOFHE_P6      0.6 cm from CEHS10      0.49814      0.00033      0.49880 CEHS17      HS_NOFHE_P7      0.7 cm from CEHS10      0.49383      0.00034      0.49451 6-225
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                          Rev. 16, May 2021 Case CEHS10, base compressed            Case CEHS13, 0.3 cm pitch expansion Case CEHS15, 0.5 cm pitch expansion      Case CEHS17, 0.7 cm pitch expansion Figure 6.13 Damaged Fuel Cases 6-226
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                Rev. 16, May 2021 6.13.6 Evaluation of Package Arrays under Normal Conditions of Transport The most reactive fuel element is determined in Section 6.13.4, Most Reactive Fuel Element Model and consists of HEU fuel meat, minimum active fuel length, minimum cladding thickness, minimum fuel meat thickness, maximum fuel meat arc length, and no fuel element spacers. A fuel element with these parameters is used in the following analysis.
6.13.6.1 NCT Array Configuration The NCT array model is a 9x9x1 lattice for a total of 81 packages, see Figure 6.13-7. Axial stacking configurations would lower the reactivity because the package is long and narrow, and the fuel element active length is significantly shorter than the package length. The geometry of the individual packages is the same as the NCT single package model. The entire array is reflected with 12-in of full-density water.
It is conservatively assumed that water may enter the package cavity. Within the cavity there are three distinct regions that may contain water:
: 1. Water within the fuel element boundary.
: 2. Water between the fuel element outer boundary and the inner diameter of the CFHE
: 3. Water between the CFHE and the inner diameter of the package cavity.
The density of water is independently varied in each of these three regions to determine the most reactive condition.
Results are provided in Table 6.13-22. The parametric study to determine the most reactive fuel element parameters (see Section 6.13.4, Most Reactive Fuel Element Model) is based upon a fully flooded NCT array model. The most reactive case from the parameter study, Case CEP11, is used as the base case for the NCT array analysis. Case CEP11 features a CFHE of nominal thickness (0.19-in). Case CENA1 features a CFHE with maximum thickness (0.25-in), while Case CENA2 features a CFHE with minimum thickness (0.13-in). For a package array, reactivity is maximized for a maximum thickness CFHE because aluminum does not appreciably absorb neutrons, and neutron transmission between packages is enhanced for the thicker CFHE.
Therefore, all subsequent NCT array cases feature the CFHE modeled at maximum thickness.
In Series 1, the water densities of the three regions are modeled at the same value and varied between 0.5 g/cm3 and 1.0 g/cm3. Reactivity is maximized for Case CENA1 (ks = 0.68564) with full water density in all model regions because maximizing the water density in the fuel element maximizes moderation.
In Series 2, the water density within the fuel element is fixed at 1.0 g/cm3, while the water density between the fuel element and the CFHE and between the CFHE and package inner diameter is modeled at the same value and varied between 0 g/cm3 and 1.0 g/cm3. The maximum reactivity of Series 2 increases compared to Series 1, with a maximum ks = 0.70962 (Case CENA15). In this case, the water density of the two outer regions is fixed at 0.5 g/cm3.
In Series 3, the water density within the fuel element and between the fuel element and CFHE is fixed at 1.0 g/cm3 while the water density between the CFHE and package inner diameter is varied. The maximum reactivity of Series 3 increases compared to Series 2, with a maximum ks 6-227
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                Rev. 16, May 2021
= 0.73887 (Case CENA20). In this case, void is modeled between the CFHE and package inner diameter.
Because Series 3 indicates maximum reactivity is achieved with void in the outer region, it is possible reactivity could be increased further by reducing the water density between the fuel element and CFHE. Series 4 is essentially the same as Series 3 except the water density between the fuel element and CFHE is fixed at a reduced value of 0.9 g/cm3. However, the most reactive Series 4 case is less reactive than Series 3, with ks = 0.73354 (Case CENA30).
The most reactive NCT array case is Case CENA20, with full-density water within the fuel element and CFHE and void between the CFHE and package inner diameter. For this case, ks =
0.73887, which is below the USL of 0.9209. This case meets the requirements of 71.55(d) and 71.59.
6.13.6.2 NCT Array Results Following are the tabulated results for the NCT array cases. The most reactive configurations are listed in boldface.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 Table 6.13 NCT Array Results Water Density (g/cm3)
Fuel      Fuel Element    CFHE to                              ks Case ID        Filename          Element      to CFHE      Package        keff            (k+2)
Series 1: Variable water density in all regions CEP11    NA_GMAX                    1.0          1.0            1.0      0.68369  0.00038    0.68445 CENA1    NA_W100_FHEMAX              1.0          1.0            1.0      0.68486  0.00039    0.68564 CENA2    NA_W100_FHEMIN              1.0          1.0            1.0      0.68043  0.00038    0.68119 CENA3    NA_W050                    0.5          0.5            0.5      0.58506  0.00037    0.58580 CENA4    NA_W060                    0.6          0.6            0.6      0.61932  0.00038    0.62008 CENA5    NA_W070                    0.7          0.7            0.7      0.64264  0.00037    0.64338 CENA6    NA_W080                    0.8          0.8            0.8      0.66049  0.00039    0.66127 CENA7    NA_W090                    0.9          0.9            0.9      0.67470  0.00039    0.67548 Series 2: Full-density water in fuel element, variable water density in remaining regions.
CENA10    NA_I100_O000                1.0            0              0        0.65080  0.00037    0.65154 CENA11    NA_I100_O010                1.0          0.1            0.1      0.67469  0.00039    0.67547 CENA12    NA_I100_O020                1.0          0.2            0.2      0.69185  0.00039    0.69263 CENA13    NA_I100_O030                1.0          0.3            0.3      0.70272  0.00039    0.70350 CENA14    NA_I100_O040                1.0          0.4            0.4      0.70609  0.00040    0.70689 CENA15    NA_I100_O050                1.0          0.5            0.5      0.70882  0.00040    0.70962 CENA16    NA_I100_O060                1.0          0.6            0.6      0.70762  0.00041    0.70844 CENA17    NA_I100_O070                1.0          0.7            0.7      0.70293  0.00038    0.70369 CENA18    NA_I100_O080                1.0          0.8            0.8      0.69780  0.00041    0.69862 CENA19    NA_I100_O090                1.0          0.9            0.9      0.69169  0.00039    0.69247 CENA1    NA_W100_FHEMAX              1.0          1.0            1.0      0.68486  0.00039    0.68564 Series 3: Full-density water in fuel element and CFHE, variable water density CFHE to package ID.
CENA20    NA_F100_O000                1.0          1.0            0        0.73807  0.00040    0.73887 CENA21    NA_F100_O010                1.0          1.0            0.1      0.73353  0.00041    0.73435 CENA22    NA_F100_O020                1.0          1.0            0.2      0.73097  0.00041    0.73179 CENA23    NA_F100_O030                1.0          1.0            0.3      0.72617  0.00039    0.72695 CENA24    NA_F100_O040                1.0          1.0            0.4      0.72085  0.00041    0.72167 CENA25    NA_F100_O050                1.0          1.0            0.5      0.71520  0.00039    0.71598 CENA26    NA_F100_O060                1.0          1.0            0.6      0.70969  0.00040    0.71049 CENA27    NA_F100_O070                1.0          1.0            0.7      0.70467  0.00042    0.70551 CENA28    NA_F100_O080                1.0          1.0            0.8      0.69780  0.00041    0.69862 CENA29    NA_F100_O090                1.0          1.0            0.9      0.69106  0.00038    0.69182 CENA1    NA_W100_FHEMAX              1.0          1.0            1.0      0.68486  0.00039    0.68564 (1/2) 6-229
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 Table 6.13 NCT Array Results (continued)
Water Density (g/cm3)
Fuel    Fuel Element      CFHE to                            ks Case ID        Filename        Element      to CFHE        Package      keff            (k+2)
Series 4: Full-density water in fuel element, 0.9 g/cm3 water in CFHE, variable water density CFHE to package ID.
CENA30    NA_F090_O000            1.0          0.9            0      0.73274  0.00040    0.73354 CENA31    NA_F090_O010            1.0          0.9            0.1      0.73111  0.00042    0.73195 CENA32    NA_F090_O020            1.0          0.9            0.2      0.72801  0.00038    0.72877 CENA33    NA_F090_O030            1.0          0.9            0.3      0.72470  0.00039    0.72548 CENA34    NA_F090_O040            1.0          0.9            0.4      0.71835  0.00039    0.71913 CENA35    NA_F090_O050            1.0          0.9            0.5      0.71497  0.00040    0.71577 CENA36    NA_F090_O060            1.0          0.9            0.6      0.70922  0.00040    0.71002 CENA37    NA_F090_O070            1.0          0.9            0.7      0.70390  0.00040    0.70470 CENA38    NA_F090_O080            1.0          0.9            0.8      0.69831  0.00039    0.69909 CENA19    NA_I100_O090            1.0          0.9            0.9      0.69169  0.00039    0.69247 (2/2) 6-230
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                        Rev. 16, May 2021 Figure 6.13 NCT Array Geometry 6-231
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 6.13.7 Package Arrays under Hypothetical Accident Conditions The most reactive fuel element is determined in Section 6.13.4, Most Reactive Fuel Element Model and consists of HEU fuel meat, minimum active fuel length, minimum cladding thickness, minimum fuel meat thickness, maximum fuel meat arc length, and no fuel element spacers. A fuel element with these parameters is used in the following analysis.
6.13.7.1 HAC Array Configuration The HAC array model is a 5x5x1 array of the HAC single package model. Initially it is assumed that the fuel element and CFHE are undamaged in the accident, as shown in Figure 6.13-8.
Results for undamaged fuel and CFHE are provided in Table 6.13-23. If the fuel element and CFHE are undamaged, the configuration is similar to the NCT array but with a smaller array size. Based on the NCT array results, an array configuration with undamaged fuel is most reactive with full-density water within the fuel element and between the fuel element and CFHE and variable density water between the CFHE and package inner diameter. In Series 1, these moderation conditions are considered, and the insulation/void in the outer region is modeled consistent with undamaged geometry. Case CEHA4 is most reactive, with 0.3 g/cm3 water between the CFHE and package inner diameter and ks = 0.64964.
For HAC, it is assumed the region between the insulation and the outer boundary of the package may contain variable density water. In Series 2, it is demonstrated that adding water to this region reduces the reactivity because neutron interactions between packages are reduced. In Series 3, it is further assumed that the insulation itself may be replaced with void or variable density water. The system is slightly more reactive when the insulation is present, consistent with the results of the ATR fuel element analysis (see Section 6.6, Package Arrays under Hypothetical Accident Conditions).
In the next three series of cases, it is assumed the fuel element and CFHE are damaged. Fuel element damage is modeled as pitch expansion in the same manner as the single package cases, and the CFHE is omitted from the models. Results for the damaged fuel element cases are provided in Table 6.13-24. In Series 4, the pitch is varied from compressed to expanded 0.7 cm relative to the compressed case. The fuel element is centered within the package cavity, as illustrated in Figure 6.13-9, and full-density water is modeled within the fuel element and within the cavity. Case CEHA46 is the most reactive, with ks = 0.75031 and 0.6 cm pitch expansion.
In Series 5, the fuel elements are shifted to the center of the array, which is a more compact geometry, see Figure 6.13-10. Reactivity increases due to the more compact geometry, with ks =
0.76369 for Case CEHA55 with 0.5 cm pitch expansion.
In Series 6, the 0.5 cm pitch expansion case from Series 5 is used as the base case and the water between the fuel element and the package inner diameter is varied between 0 and 1.0 g/cm3. A small increase in reactivity is observed for Case CEHA68, with 0.8 g/cm3 water in this region and ks = 0.76431.
Therefore, Case CEHA68 is the most reactive HAC array case with damaged fuel, with ks =
0.76431, which is below the USL of 0.9209. This case features pitch expansion of 0.5 cm, fully moderated fuel elements, and 0.8 g/cm3 water between the fuel element and package cavity.
This case meets the requirements of 71.55(e) and 71.59.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                              Rev. 16, May 2021 6.13.7.2 HAC Array Results Following are the tabulated results for the HAC array cases. The most reactive configurations are listed in boldface.
Table 6.13 HAC Array Results, Undamaged Fuel Water Density (g/cm3)
CFHE to        Insulation      Insulation                              ks Case ID          Filename            Package          Region        to Package          keff            (k+2)
Series 1: Water density of 1.0 g/cm3 within fuel element and between fuel element and CFHE. Variable water density between CFHE and package ID, insulation modeled, void between insulation and package wall.
CEHA1      HA_FHE_O000                    0          Insulation            0          0.64160    0.00039  0.64238 CEHA2      HA_FHE_O010                    0.1          Insulation            0          0.64511    0.00040  0.64591 CEHA3      HA_FHE_O020                    0.2          Insulation            0          0.64730    0.00039  0.64808 CEHA4      HA_FHE_O030                    0.3          Insulation            0          0.64888    0.00038  0.64964 CEHA5      HA_FHE_O040                    0.4          Insulation            0          0.64870    0.00041  0.64952 CEHA6      HA_FHE_O050                    0.5          Insulation            0          0.64778    0.00040  0.64858 CEHA7      HA_FHE_O060                    0.6          Insulation            0          0.64649    0.00042  0.64733 CEHA8      HA_FHE_O070                    0.7          Insulation            0          0.64407    0.00040  0.64487 CEHA9      HA_FHE_O080                    0.8          Insulation            0          0.64226    0.00040  0.64306 CEHA10    HA_FHE_O090                    0.9          Insulation            0          0.63958    0.00037  0.64032 CEHA11    HA_FHE_O100                    1.0          Insulation            0          0.63536    0.00039  0.63614 3
Series 2: Water density of 1.0 g/cm within fuel element and between fuel element and CFHE. Water density of 0.3 g/cm3 between CFHE and package ID, insulation modeled, variable density water between insulation and package wall.
CEHA4      HA_FHE_O030                    0.3          Insulation            0          0.64888    0.00038  0.64964 CEHA20    HA_FHE_O030_Z010              0.3          Insulation          0.1        0.64299    0.00040  0.64379 CEHA21    HA_FHE_O030_Z020              0.3          Insulation          0.2        0.63703    0.00038  0.63779 CEHA22    HA_FHE_O030_Z030              0.3          Insulation          0.3        0.62913    0.00039  0.62991 3
Series 3: Water density of 1.0 g/cm within fuel element and between fuel element and CFHE. Water density of 0.3 g/cm3 between CFHE and package ID. Variable density water between inner and outer tubes.
CEHA30    HA_FHE_O030_ZZ000              0.3              0                0          0.64554    0.00040  0.64634 CEHA31    HA_FHE_O030_ZZ010              0.3            0.1              0.1        0.62597    0.00039  0.62675 CEHA32    HA_FHE_O030_ZZ020              0.3            0.2              0.2        0.60422    0.00037  0.60496 CEHA33    HA_FHE_O030_ZZ030              0.3            0.3              0.3        0.58257    0.00038  0.58333 When the insulation is modeled as insulation rather than water or void, insulation is denoted.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                Rev. 16, May 2021 Table 6.13 HAC Array Results, Damaged Fuel Water Density, Fuel Element to ID                              ks Case ID      Filename          Change in Pitch        (g/cm3)        keff                (k+2)
Series 4: Damaged fuel with variable pitch, no CFHE, full-density water within cavity, fuel elements centered within package.
CEHA40      HA_P0            Base Compressed              1.0        0.60171    0.00039      0.60249 CEHA41      HA_P1            0.1 cm from CEHA40          1.0        0.64885    0.00039      0.64963 CEHA42      HA_P2            0.2 cm from CEHA40          1.0        0.68612    0.00040      0.68692 CEHA43      HA_P3            0.3 cm from CEHA40          1.0        0.71493    0.00040      0.71573 CEHA44      HA_P4            0.4 cm from CEHA40          1.0        0.73454    0.00038      0.73530 CEHA45      HA_P5            0.5 cm from CEHA40          1.0        0.74604    0.00038      0.74680 CEHA46 HA_P6                  0.6 cm from CEHA40          1.0        0.74959    0.00036      0.75031 CEHA47      HA_P7            0.7 cm from CEHA40          1.0        0.74407    0.00034      0.74475 Series 5: Damaged fuel with variable pitch, no CFHE, full-density water within cavity, fuel elements shifted to center of array.
CEHA50      HA_CP0            Base Compressed              1.0        0.63734    0.00039      0.63812 CEHA51      HA_CP1            0.1 cm from CEHA40          1.0        0.68523    0.00038      0.68599 CEHA52      HA_CP2            0.2 cm from CEHA40          1.0        0.72004    0.00037      0.72078 CEHA53      HA_CP3            0.3 cm from CEHA40          1.0        0.74422    0.00039      0.74500 CEHA54      HA_CP4            0.4 cm from CEHA40          1.0        0.75836    0.00039      0.75914 CEHA55 HA_CP5                0.5 cm from CEHA40          1.0        0.76297    0.00036      0.76369 CEHA56      HA_CP6            0.6 cm from CEHA40          1.0        0.75998    0.00037      0.76072 CEHA57      HA_CP7            0.7 cm from CEHA40          1.0        0.74748    0.00036      0.74820 Series 5: Damaged fuel with pitch increased 0.5 cm from Case CEHA40, no CFHE, full-density water within fuel element, variable density water between fuel element and ID of package cavity, fuel elements shifted to center of array.
CEHA60      HA_CP5_O000      0.5 cm from CEHA40            0        0.73839    0.00037      0.73913 CEHA61      HA_CP5_O010      0.5 cm from CEHA40          0.1        0.74329    0.00037      0.74403 CEHA62      HA_CP5_O020      0.5 cm from CEHA40          0.2        0.74833    0.00037      0.74907 CEHA63      HA_CP5_O030      0.5 cm from CEHA40          0.3        0.75285    0.00038      0.75361 CEHA64      HA_CP5_O040      0.5 cm from CEHA40          0.4        0.75600    0.00037      0.75674 CEHA65      HA_CP5_O050      0.5 cm from CEHA40          0.5        0.75924    0.00036      0.75996 CEHA66      HA_CP5_O060      0.5 cm from CEHA40          0.6        0.76080    0.00037      0.76154 CEHA67      HA_CP5_O070      0.5 cm from CEHA40          0.7        0.76314    0.00039      0.76392 CEHA68 HA_CP5_O080            0.5 cm from CEHA40          0.8        0.76355    0.00038      0.76431 CEHA69      HA_CP5_O090      0.5 cm from CEHA40          0.9        0.76312    0.00036      0.76384 CEHA55      HA_CP5            0.5 cm from CEHA40          1.0        0.76297    0.00036      0.76369 6-234
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                      Rev. 16, May 2021 Figure 6.13 HAC Array Geometry, Undamaged Geometry 6-235
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                    Rev. 16, May 2021 Figure 6.13 HAC Array Geometry, Damaged Geometry, Centered 6-236
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                      Rev. 16, May 2021 Figure 6.13 HAC Array Geometry, Damaged Geometry, Shifted 6-237
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 6.13.8 Fissile Material Packages for Air Transport See Section 6.7, which applies to all contents.
6.13.9 Benchmark Evaluations Cobra fuel is high-enriched plate-type fuel, similar to ATR fuel. Therefore, the benchmarking evaluation performed for ATR fuel documented in Section 6.8, Benchmark Evaluations, is applicable to the current analysis, and the USL is 0.9209.
Five parameters were selected for the benchmark evaluation: (1) energy of the average neutron lethargy causing fission (EALF), (2) U-235 number density, (3) channel width, (4) H/U-235 atom ratio, and (5) pitch. The range of applicability of these parameters for the benchmarks utilized is summarized in Table 6.8-2. In the following sections, the range of applicability of the benchmarks is compared with the Cobra criticality analysis.
EALF The most reactive HAC array case CEHA68 has an EALF = 4.8847x10-8 MeV, which is slightly outside the minimum bound of 5.22210x10-8 MeV of the benchmark experiments. Because this parameter is only slightly outside the bounds of the benchmark experiments, extrapolation of the USL is acceptable. The USL equation for EALF based on the 17 experiment benchmark set is USL = 0.9171 + (7.9546E+04)*EALF, or USL = 0.9210 for an EALF = 4.8847x10-8 MeV. A USL of 0.9210 is larger than the USL of 0.9209 used in the analysis; thus, a USL of 0.9209 is acceptable.
In addition, the small quantity payload criticality analysis utilizes HEU solution benchmarks with a minimum EALF of 3.43x10-8 MeV (see Section 6.11.8, Benchmark Evaluations), which bounds the EALF of the Cobra analysis. The solution benchmarks resulted in a USL of 0.9309; therefore, adding solution benchmarks to the flat-plate benchmark set would increase the USL.
Therefore, it is conservative to apply a USL of 0.9209 for the Cobra criticality analysis, and the EALF parameter is acceptable.
U-235 Number Density The U-235 number density of the HEU cases is approximately 2.7x10-3 atom/b-cm. This is within the range of applicability of 1.849x10-3  X  3.926x10-3 atom/b-cm.
Channel Width, H/U-235 Atom Ratio, and Pitch The channel width (i.e., water gap between plates), H/U-235 atom ratio, and plate pitch are all related to the system moderation conditions and are somewhat redundant. The EALF parameter is a better metric for determining system moderation than channel width, H/U-235 atom ratio, or plate pitch because it is directly related to the energy range causing fission. Due to the damaged fuel assumption in the analysis and the large channel widths and pitches used in the models, benchmarks with similarly large parameters for a plate fuel element are not available. Because the EALF parameter is shown above to be acceptable for the Cobra fuel analysis, the channel width, H/U-235 atom ratio, and plate pitch parameters are also acceptable.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                            Rev. 16, May 2021 6.13.10      Sample Input File A sample input file for the most reactive case (Case CEHA68, HAC array, filename HA_CP5_O080) is included below.
ATR 999      0        -320:321:-322:323:-324:325                      imp:n=0 900      0        310 -311 312 -313 24 -25            fill=3      imp:n=1 901      2 -1.0 (311:-310:313:-312:-24:25) 320 -321 322 -323 324 -325                imp:n=1 c
c        Universe 1: Fuel Element in FHE c
101      0        101  -102  110  -111  u=1  fill=20  imp:n=1 102      0        101  -102  111  -112  u=1  fill=21  imp:n=1 103      0        101  -102  112  -113  u=1  fill=22  imp:n=1 104      0        101  -102  113  -114  u=1  fill=23  imp:n=1 105      0        101  -102  114  -115  u=1  fill=24  imp:n=1 106      0        101  -102  115  -116  u=1  fill=25  imp:n=1 c
111      0        -101    100  110  -111  u=1  fill=20(1) imp:n=1 112      0        -101    100  111  -112  u=1  fill=21(1) imp:n=1 113      0        -101    100  112  -113  u=1  fill=22(1) imp:n=1 114      0        -101    100  113  -114  u=1  fill=23(1) imp:n=1 115      0        -101    100  114  -115  u=1  fill=24(1) imp:n=1 116      0        -101    100  115  -116  u=1  fill=25(1) imp:n=1 c
121      0        102  -100  110  -111  u=1  fill=20(2)  imp:n=1 122      0        102  -100  111  -112  u=1  fill=21(2)  imp:n=1 123      0        102  -100  112  -113  u=1  fill=22(2)  imp:n=1 124      0        102  -100  113  -114  u=1  fill=23(2)  imp:n=1 125      0        102  -100  114  -115  u=1  fill=24(2)  imp:n=1 126      0        102  -100  115  -116  u=1  fill=25(2)  imp:n=1 c
130      2 -1.0 -110                      u=1              imp:n=1 $ inner water 131      2 -0.8 116                        u=1              imp:n=1 $ outer water c
c        Universe 2: Package, fuel in center c
201      2 -1.0              -200 fill=1                u=2 imp:n=1 202      4 -7.94          200 -201                        u=2 imp:n=1 $ pipe 203      6 -0.096        201 -203 250 -251 252 -253 u=2      imp:n=1 $ insulation 204      0                203 250 -251 252 -253          u=2 imp:n=1 $ insulation to tube 205      4 -7.94          -250:251:-252:253              u=2 imp:n=1 $ tube to inf c
c        Universe 30: Package, fuel shifted left c
3001    2 -1.0              -200 fill=1(30)            u=30 imp:n=1 3002    4 -7.94          200 -201                        u=30 imp:n=1 $ pipe 3003    6 -0.096        201 -203 250 -251 252 -253 u=30      imp:n=1 $ insulation 3004    0                203 250 -251 252 -253          u=30 imp:n=1 $ insulation to tube 3005    4 -7.94          -250:251:-252:253              u=30 imp:n=1 $ tube to inf c
c        Universe 31: Package, fuel shifted right c
3101    2 -1.0              -200    fill=1(31)          u=31 imp:n=1 6-239
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                              Rev. 16, May 2021 3102  4 -7.94      200 -201                  u=31 imp:n=1 $ pipe 3103  6 -0.096    201 -203 250 -251 252 -253 u=31 imp:n=1 $ insulation 3104  0            203 250 -251 252 -253      u=31 imp:n=1 $ insulation to tube 3105  4 -7.94      -250:251:-252:253          u=31 imp:n=1 $ tube to inf c
c      Universe 32: Package, fuel shifted down c
3201  2 -1.0            -200 fill=1(32)      u=32 imp:n=1 3202  4 -7.94      200 -201                  u=32 imp:n=1 $ pipe 3203  6 -0.096    201 -203 250 -251 252 -253 u=32 imp:n=1 $ insulation 3204  0            203 250 -251 252 -253      u=32 imp:n=1 $ insulation to tube 3205  4 -7.94      -250:251:-252:253          u=32 imp:n=1 $ tube to inf c
c      Universe 33: Package, fuel shifted up c
3301  2 -1.0            -200 fill=1(33)      u=33 imp:n=1 3302  4 -7.94      200 -201                  u=33 imp:n=1 $ pipe 3303  6 -0.096    201 -203 250 -251 252 -253 u=33 imp:n=1 $ insulation 3304  0            203 250 -251 252 -253      u=33 imp:n=1 $ insulation to tube 3305  4 -7.94      -250:251:-252:253          u=33 imp:n=1 $ tube to inf c
c      Universe 34: Package, fuel shifted lower left c
3401  2 -1.0            -200 fill=1(34)      u=34 imp:n=1 3402  4 -7.94      200 -201                  u=34 imp:n=1 $ pipe 3403  6 -0.096    201 -203 250 -251 252 -253 u=34 imp:n=1 $ insulation 3404  0            203 250 -251 252 -253      u=34 imp:n=1 $ insulation to tube 3405  4 -7.94      -250:251:-252:253          u=34 imp:n=1 $ tube to inf c
c      Universe 35: Package, fuel shifted lower right c
3501  2 -1.0            -200 fill=1(35)      u=35 imp:n=1 3502  4 -7.94      200 -201                  u=35 imp:n=1 $ pipe 3503  6 -0.096    201 -203 250 -251 252 -253 u=35 imp:n=1 $ insulation 3504  0            203 250 -251 252 -253      u=35 imp:n=1 $ insulation to tube 3505  4 -7.94      -250:251:-252:253          u=35 imp:n=1 $ tube to inf c
c      Universe 36: Package, fuel shifted upper left c
3601  2 -1.0            -200 fill=1(36)      u=36 imp:n=1 3602  4 -7.94      200 -201                  u=36 imp:n=1 $ pipe 3603  6 -0.096    201 -203 250 -251 252 -253 u=36 imp:n=1 $ insulation 3604  0            203 250 -251 252 -253      u=36 imp:n=1 $ insulation to tube 3605  4 -7.94      -250:251:-252:253          u=36 imp:n=1 $ tube to inf c
c      Universe 37: Package, fuel shifted upper right c
3701  2 -1.0            -200 fill=1(37)      u=37 imp:n=1 3702  4 -7.94      200 -201                  u=37 imp:n=1 $ pipe 3703  6 -0.096    201 -203 250 -251 252 -253 u=37 imp:n=1 $ insulation 6-240
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                Rev. 16, May 2021 3704    0          203 250 -251 252 -253      u=37 imp:n=1 $ insulation to tube 3705    4 -7.94    -250:251:-252:253          u=37 imp:n=1 $ tube to inf c
c      Universe 3: Array of Packages c
300  0    -300 301 -302 303 imp:n=1 u=3 lat=1 fill=-2:2 -2:2 0:0 36 36 33 37 37 36 36 33 37 37 31 31 2 30 30 35 35 32 34 34 35 35 32 34 34 c
c      Universe 10: Wedge 1 c
1000  3  -2.7    1000 -1010 1020 -1030  u=10 imp:n=1 1001  10 -3.3532 1000 -1010 1030 -1040  u=10 imp:n=1 1002  3  -2.7    1000 -1010 1040 -1050  u=10 imp:n=1 1003  2  -1.0  -1000:1010:-1020:1050  u=10 imp:n=1 c
c      Universe 11: Wedge 2 c
1100  3  -2.7    1001 -1011 1021 -1031  u=11 imp:n=1 1101  10 -3.3995 1001 -1011 1031 -1041  u=11 imp:n=1 1102  3  -2.7    1001 -1011 1041 -1051  u=11 imp:n=1 1103  2  -1.0  -1001:1011:-1021:1051  u=11 imp:n=1 c
c      Universe 12: Wedge 3 c
1200  3  -2.7    1002 -1012 1022 -1032  u=12 imp:n=1 1201  10 -3.4569 1002 -1012 1032 -1042  u=12 imp:n=1 1202  3  -2.7    1002 -1012 1042 -1052  u=12 imp:n=1 1203  2  -1.0  -1002:1012:-1022:1052  u=12 imp:n=1 c
c      Universe 13: Wedge 4 c
1300  3  -2.7    1003 -1013 1023 -1033  u=13 imp:n=1 1301  10 -3.4540 1003 -1013 1033 -1043  u=13 imp:n=1 1302  3  -2.7    1003 -1013 1043 -1053  u=13 imp:n=1 1303  2  -1.0  -1003:1013:-1023:1053  u=13 imp:n=1 c
c      Universe 14: Wedge 5 c
1400  3  -2.7    1004 -1014 1024 -1034  u=14 imp:n=1 1401  10 -3.5112 1004 -1014 1034 -1044  u=14 imp:n=1 1402  3  -2.7    1004 -1014 1044 -1054  u=14 imp:n=1 1403  2  -1.0  -1004:1014:-1024:1054  u=14 imp:n=1 c
c      Universe 15: Wedge 6 c
1500  3  -2.7    1005 -1015 1025 -1035  u=15 imp:n=1 1501  10 -3.4735 1005 -1015 1035 -1045  u=15 imp:n=1 1502  3  -2.7    1005 -1015 1045 -1055  u=15 imp:n=1 1503  2  -1.0  -1005:1015:-1025:1055  u=15 imp:n=1 c
c    Universe 20, 21, 22, 23, 24, 25, pitch changes c
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                        Rev. 16, May 2021 2000  0        -2000  fill=10(20) u=20  imp:n=1 2001  0        -2000  fill=11(21) u=21  imp:n=1 2002  0        -2000  fill=12(22) u=22  imp:n=1 2003  0        -2000  fill=13(23) u=23  imp:n=1 2004  0        -2000  fill=14(24) u=24  imp:n=1 2005  0        -2000  fill=15(25) u=25  imp:n=1 24      pz -37.475                  $ bottom of fuel 25      pz 37.475                    $ top of fuel c
100    py 0 101    p 1.7321 -1 0 0 102    p -1.7321 -1 0 0 110  cz  0.2    $inner bound 111  cz  1.5305 $mid-112  cz  2.463 $mid-113  cz  3.396 $mid-114  cz  4.3285 $mid-115  cz  5.261 $mid-116  cz  6.2 $outer bound-120  cz  5.207  $FHE inner 121  cz  5.6896 $FHE outer c
200      cz 7.3838 $ IR pipe 201      cz 7.6581 $ OR pipe 203      cz 10.1981 $ 1" insulation 204      pz 1000    $ dummy c
250      px  -9.6032 $ square tube 251      px  9.6032 252      py  -9.6032 253      py  9.6032 c
300      px  10.033 $ lattice surfaces/sq. tube 301      px -10.033 302      py  10.033 303      py -10.033 310      px -50.164 $ 5x5 bounds 311      px  50.164 312      py -50.164 313      py  50.164 320      px -80.644 $ outer bounds 321      px  80.644 322      py -80.644 323      py  80.644 324      pz -67.955 325      pz  67.955 c
1000  p  1.1181  -1  0  0  $upper 1001  p  1.2186  -1  0  0  $upper 1002  p  1.2923  -1  0  0  $upper 1003  p  1.3479  -1  0  0  $upper 1004  p  1.3921  -1  0  0  $upper 1005  p  1.4249  -1  0  0  $upper 1010  p  -1.1181  -1  0  0  $lower 1011  p  -1.2186  -1  0  0  $lower 1012  p  -1.2923  -1  0  0  $lower 6-242
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                      Rev. 16, May 2021 1013  p -1.3479  -1 0    0  $lower 1014  p -1.3921  -1 0    0  $lower 1015  p -1.4249  -1 0    0  $lower 1020  cz 1.6055  $ r1 1021  cz 2.0375  $ r1 1022  cz 2.4705  $ r1 1023  cz 2.9035  $ r1 1024  cz 3.3355  $ r1 1025  cz 3.7685  $ r1 1030  cz 1.6355  $r2 1031  cz 2.0675  $r2 1032  cz 2.5005  $r2 1033  cz 2.9335  $r2 1034  cz 3.3655  $r2 1035  cz 3.7985  $r2 1040  cz 1.6935  $r3 1041  cz 2.1255  $r3 1042  cz 2.5585  $r3 1043  cz 2.9915  $r3 1044  cz 3.4235  $r3 1045  cz 3.8565  $r3 1050  cz 1.7235  $r4 1051  cz 2.1555  $r4 1052  cz 2.5885  $r4 1053  cz 3.0215  $r4 1054  cz 3.4535  $r4 1055  cz 3.8865  $r4 c
2000  so 10000 m2      1001.62c 2            $ water 8016.62c 1 mt2    lwtr.60t m3      13027.62c 1          $ Al m4      6000.66c  -0.08      $ SS-304 14000.60c -1.0 15031.66c -0.045 24000.50c -19.0 25055.62c -2.0 26000.55c -68.375 28000.50c -9.5 m6      13027.62c -26.5      $ insulation material 14000.60c -23.4 8016.62c  -50.2 m10    92235.69c -3.0136E-01    $meat 92238.69c -1.9236E-02 13027.62c -6.7941E-01 c
*tr1    0 0 0 120 30 90 150 120 90 $ 120 deg CCW
*tr2    0 0 0 120 150 90 30 120 90 $ 120 deg CW
*tr20 -0.35 0 0
*tr21 0.15 0 0
*tr22 0.65 0 0
*tr23 1.15 0 0
*tr24 1.65 0 0
*tr25 2.15 0 0 tr30 -1.1828 0 0 $ shift left 6-243
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                    Rev. 16, May 2021 tr31  1.1828 0 0 $ shift right tr32  0 -1.1828 0 $ shift down tr33  0 1.1828 0 $ shift up tr34  -0.8364 -0.8364 0 $ fuel shifted lower left tr35  0.8364 -0.8364 0 $ fuel shifted lower right tr36  0.8364 0.8364 0 $ fuel shifted upper left tr37  -0.8364 0.8364 0 $ fuel shifted upper right c
mode  n kcode  5000 1.0 50 850 sdef    x=d1 y=d2 z=d3 si1    -50 50 sp1    0 1 si2    -50 50 sp2    0 1 si3    -40 40 sp3    0 1 6-244
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                    Rev. 16, May 2021 6.14 Appendix F: Criticality Analysis for ATR, MURR, MIT, and NBSR LEU Fuel Elements and/or DDEs The ATR FFSC may be utilized to transport LEU versions of the ATR, MURR, and MIT fuel elements analyzed previously in this chapter. These fuel elements are plate-type fuels with similar geometry and layout to the previously-discussed HEU versions. The ATR FFSC may also be utilized to transport LEU DDEs for the MURR and MIT. DDEs have minor dimensional variations from standard fuel elements to support design verification and testing but are otherwise very similar to the LEU fuel elements.
Additionally, the ATR FFSC may be utilized to transport the LEU DDE for the NBSR. Similar to the other LEU payloads, this DDE is a plate-type fuel.
Due to significant similarities in materials, construction, and criticality behavior, the aforementioned LEU fuel elements and DDEs are jointly analyzed within this section. The following analyses demonstrate that the ATR FFSC when transporting ATR, MURR, MIT, and NBSR LEU fuel elements and/or DDEs complies with the requirements of 10 CFR 71.55 and 71.59. The Criticality Safety Index (CSI), per 10 CFR 71.59, is 6.25 for the ATR LEU fuel element and 4.0 for the MURR, MIT, and NBSR LEU fuel elements and/or DDEs.
6.14.1 Description of Criticality Design 6.14.1.1 Design Features No special design features are required to maintain criticality safety. No poisons are utilized in the package. The separation provided by the packaging (outer flat-to-flat dimension of 7.9-in),
along with the limit on the number of packages per shipment, is sufficient to maintain criticality safety.
6.14.1.2 Summary Table of Criticality Evaluation The USL for ATR, MURR, MIT, and NBSR LEU fuel elements and/or DDEs is calculated on a case-by-case basis in Section 6.14.6, Benchmark Evaluations. The package is considered to be acceptably subcritical if the computed ksafe (ks), which is defined as keffective (keff) plus twice the statistical uncertainty (), is less than or equal to the USL, or:
ks = keff + 2  USL The USL is determined on the basis of a benchmark analysis and incorporates the combined effects of code computational bias, the uncertainty in the bias based on both benchmark-model and computational uncertainties, and an administrative margin. The results of the benchmark analysis indicate that the USL is adequate to ensure subcriticality of the package.
The packaging design is shown to meet the requirements of 10 CFR 71.55(b)(d)(e) [8] (Note:
references for this section are found in Section 6.14.7, References). Moderation by water in the most reactive credible extent is utilized in both the normal conditions of transport (NCT) and hypothetical accident conditions (HAC) analyses. In the single package NCT models, full-density water fills the accessible cavity, while in the single package HAC models, full-density water fills all cavities. In all single package and array models, 12-in of water reflection is utilized. In the NCT and HAC array cases, partial moderation is considered to maximize array interaction effects.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                              Rev. 16, May 2021 The maximum results of the criticality calculations are summarized in Table 6.14-1.
Table 6.14 Summary of Criticality Evaluation (LEU Fuel Elements and DDEs)
ATR LEU        MURR LEU      MIT LEU      NBSR LEU Normal Conditions of Transport (NCT)
Case                  ks              ks            ks            ks Single Unit Maximum        0.42385          0.5063      0.42693        0.39137 0.77809        0.86146      0.68613        0.65257 Array Maximum (7x7)            (8x8)        (9x9)          (9x9)
Hypothetical Accident Conditions (HAC)
Case                  ks              ks            ks            ks Single Unit Maximum        0.58853        0.58572      0.52045        0.50968 0.87986        0.86625      0.72735        0.74897 Array Maximum (4x4)            (5x5)        (5x5)          (5x5)
Upper Subcritical Limit (USL) 0.92312        0.92071      0.92288        0.92348 Air Transport Analysis ks = 0.71171                USL = 0.90868 6.14.1.3 Criticality Safety Index The CSI, as defined in 10 CFR 71.59, is calculated as follows:
50              #                  #
5                                  2 For the ATR LEU fuel element, the NCT and HAC arrays have 49 and 16 packages respectively, resulting in N values of 9.8 and 8. The CSI, calculated with NHAC, is 6.25. A maximum of 8 packages (nonexclusive use) or 16 packages (exclusive use) are allowed per shipment.
For the MURR, MIT, and NBSR LEU fuel elements and/or DDEs, the NCT and HAC arrays have  64 and 25 packages respectively, resulting in N values of  12.8 and 12.5. The CSI, calculated using NHAC, is 4. A maximum of 12 packages (nonexclusive use) or 25 packages (exclusive use) are allowed per shipment.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 6.14.2 Fissile Material Contents All of the LEU fuel elements and DDEs consist of multiple fuel plates slotted into opposing side plates, with the opposing side plates restricting the movement of the fuel plates relative to each other. Each fuel plate consists of aluminum cladding surrounding a centrally-located fuel foil or meat. The fuel meat and cladding are separated by 0.001-inch thick layers of zirconium (all dimensions nominal unless otherwise stated).
All fuel meat is U-10Mo. The molybdenum content is 10 +/- 1 wt.%, with impurities limited to 1,200 ppm. The remaining alloy content is uranium, with a U-235 enrichment of 19.75 +/- 0.20 wt.%. The uranium isotopic composition is shown in Table 6.14-2. The fuel meat density and dimensions are different for each fuel type.
Table 6.14 LEU U-10Mo Uranium Composition Isotope          Units              Limit U-232          &#xb5;g/gU              0.002 U-234          wt.% U            0.260%
U-235          wt.% U        19.75 +/- 0.20%
U-236          &#xb5;g/gU              4600 U-238                Remaining mass 6.14.2.1 ATR LEU Fuel Element The ATR LEU fuel element assembly consists of 19 curved fuel plates. Each ATR LEU fuel plate contains a 0.008-inch to 0.016-inch thick fuel meat. The total thickness of each plate is 0.050 inches to 0.100 inches. Fuel plate widths (by arc length) vary significantly and are between 2 inches and 4 inches, with cladding side widths between 0.160 inches to 0.190 inches.
Fuel plate #1 is the smallest plate by midplane radius, with the size progressively increasing until the final and largest plate, fuel plate #19. The fuel meat height is 48 inches, while the fuel plate height is 49.5 inches. The channel thickness between fuel plates is 0.078 inches.
The nominal fuel meat volume and U-235 mass for each fuel plate is shown in Table 6.14-3.
The total U-235 loading of an ATR LEU fuel element is 1,648 g +/- 2%, with a nominal U-10Mo alloy density of 16.5 g/cm3. The modeled U-235 loading is 1,699.2 g as shown in Table 6.14-9. The bounding value for transport is 1,681 g as shown in Table 1.1-1.
6.14.2.2 MURR LEU Fuel Element and DDE The MURR LEU fuel element assembly consists of 23 curved fuel plates. Each fuel plate contains a 0.009-inch to 0.020-inch thick fuel meat. The total thickness of each plate is 0.044 inches to 0.049 inches, with cladding thicknesses between 0.0110 inches and 0.0165 inches.
Fuel plate widths (by arc length) vary significantly and are between 1.9 inches and 4.4 inches.
Fuel plate #1 is the smallest plate by midplane radius, with the size progressively increasing until the final and largest plate, fuel plate #23. The fuel meat height is 24 inches, while the fuel plate 6-247
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 height is 25.5 inches. The channel thickness between fuel plates is 0.092 inches to 0.093 inches.
The MURR DDE will use the same fuel plates as the MURR LEU fuel element.
The U-235 mass loadings for each fuel plate are shown in Table 6.14-4. The total U-235 loading of a MURR LEU fuel element is 1,507 g +/- 15.1 g, with a nominal U-10Mo alloy density of 17.0 g/cm3. The tolerance on total U-235 mass is less than the sum of the individual plate tolerances. The modeled U-235 loading is 1,664.7 g as shown in Table 6.14-11. The bounding value for transport is 1,660 g as shown in Table 1.1-1.
6.14.2.3 MIT LEU Fuel Element and DDE The MIT LEU fuel element assembly consists of 19 offset flat plates, forming a rhombus-shaped assembly. Each fuel plate contains a 0.013-inch to 0.025-inch thick fuel meat. The total thickness of each plate is 0.049 inches, with cladding thicknesses between 0.011 inches and 0.017 inches. There are three types of fuel plates: Type T (2 outermost plates in assembly), Type F (13 innermost plates), and Type Y (remaining 4 plates). The fuel meat height is 22.375 inches, while the fuel plate height is 23.00 inches. The channel thickness between fuel plates is at most 0.082 inches, with the fuel assembly average channel thickness not to exceed 0.079 inches. The MIT DDE will use the same fuel plates as the MIT LEU fuel element.
The U-235 mass loadings for each fuel plate are shown in Table 6.14-5. The total U-235 loading of a MIT LEU fuel element is 968 g +/- 9.7 g, with a nominal U-10Mo alloy density of 17.0 g/cm3.
The tolerance on total U-235 mass is less than the sum of the individual plate tolerances. The modeled U-235 loading is 1,092.45 g as shown in Table 6.14-13. The bounding value for transport is 1,070 g as shown in Table 1.1-1.
6.14.2.4 NBSR DDE The NBSR LEU DDE consists of 34 curved fuel plates. The fuel plates are axially stacked within the assembly, with a 2-inch gap separating two sets of 17 plates. Each fuel plate contains a 0.0085-inch thick fuel meat. The total thickness of each plate is 0.050 inches. The fuel meat height is up to 11.635 inches, while the fuel plate height is 13.000 inches. The fuel plate is manufactured flat, with a width of 2.79 inches, before being rolled to a curvature radius of 5.500 inches. The channel thickness between fuel plates is 0.116 inches.
The U-235 fuel loading of a single fuel plate is 11.27 g +/- 0.24 g, resulting in a total U-235 loading of 383.18 g +/- 8.16 g for a single NBSR DDE. The nominal density of the U-10Mo alloy density is 17.2 g/cm3. The modeled U-235 loading is 481 g. The bounding value for transport is 460 g as shown in Table 1.1-1.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                Rev. 16, May 2021 Table 6.14 ATR LEU Fuel Meat Nominal Parameters Plate        Fuel Meat      U-235          Plate  Fuel Meat          U-235
  #        Volume (cm3)    Mass (g)          #  Volume (cm3)      Mass (g) 1          11.15          33.5          11        35.09          105.4 2          19.26          57.9          12        36.36          109.2 3          20.29          60.9          13        37.62          113.0 4          26.24          78.8          14        38.89          116.8 5          27.50          82.6          15        40.15          120.6 6          28.77          86.4          16        33.65          101.1 7          30.03          90.2          17        21.34            64.1 8          31.30          94.0          18        21.98            66.0 9          32.56          97.8          19        22.73            68.3 10          33.83          101.6        Total    548.74          1648.1 Table 6.14 MURR LEU U-235 Mass by Fuel Plate Plate      U-235                        Plate  U-235 Tolerance (g)                        Tolerance (g)
      #      Mass (g)                        #  Mass (g) 1        18.09        2.01            13    70.78          3.54 2        25.66        2.14            14    73.32          3.67 3        36.26        2.27            15    75.86          3.79 4        47.89        2.39            16    78.41          3.92 5        50.45        2.52            17    80.95          4.05 6        52.99        2.65            18    83.49          4.17 7        55.53        2.78            19    86.03          4.30 8        58.07        2.90            20    88.59          4.43 9        60.61        3.03            21    91.15          4.56 10        63.16        3.16            22    93.71          4.69 11        65.70        3.28            23    81.87          4.82 12        68.24        3.41          Total  1507          15.1 6-249
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                            Rev. 16, May 2021 Table 6.14 MIT LEU U-235 Mass by Fuel Plate Plate #  Plate Type U-235 Mass (g) Tolerance (g) 1          T          30.02          1.05 2          Y          39.26          1.37 3          Y          39.26          1.37 4          F          57.74          2.02 5          F          57.74          2.02 6          F          57.74          2.02 7          F          57.74          2.02 8          F          57.74          2.02 9          F          57.74          2.02 10          F          57.74          2.02 11          F          57.74          2.02 12          F          57.74          2.02 13          F          57.74          2.02 14          F          57.74          2.02 15          F          57.74          2.02 16          F          57.74          2.02 17          Y          39.26          1.37 18          Y          39.26          1.37 19          T          30.02          1.05 Total              968            9.7 6-250
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 6.14.3 General Considerations 6.14.3.1 Model Configuration All relevant design features of the LEU fuel elements, DDEs, and ATR FFSC are modeled in MCNP. Since the package is long and narrow, array configurations are stacked only in the lateral directions. All models are surrounded by 12 inches of water reflection.
Fuel element and DDE modeling is limited to the fuel plates. Non-fuel components such as end boxes and side plates are conservatively not modeled and instead replaced by variable-density moderator. For NCT, the side plates are assumed to maintain the fuel plates in a fixed spacing.
For HAC, it is assumed that the side plates could fail and thus the fuel plates can freely separate.
No damage to the fuel plates themselves is modeled. The zirconium interlayers within each fuel plate are not modeled and are instead included as part of the aluminum cladding. Fuel modeling is axially limited to the region encompassing the respective fuel meat. For conservatism, the fuel plates are modeled using worst-case tolerances.
Due to the simplifications used in modeling, separate models are not necessary to evaluate fuel elements and DDEs of the same type.
Due to its moderating qualities, neoprene material which may be present in FHEs is accounted for. Neoprene is modeled without chlorine and assumed to possibly become homogenized with water inside the ATR FFSC cavity. Neoprene is only credited when its presence is concluded to increase reactivity.
NCT and HAC single package models are shown in Figure 6.14-1 through Figure 6.14-4. The figures show the fuel meat (green, may not be visible due to size), fuel cladding (pink), flooded water (blue), insulation (yellow), and steel (brown). Empty space is modeled as void (white).
The outline of the MURR LEU FHE and MIT FHE inner cavities, with NCT dimensions as-built and HAC dimensions following damage, can be seen on the respective figures. On all figures, the fuel bag region can be seen to encompass the channels between fuel plates.
6.14.3.1.1      ATR FFSC The ATR FFSC is modeled consistent with the other criticality analyses in this chapter. The ATR FFSC model consists of the inner tube, insulation, and the outer shell. Material volumes are conservatively minimized to minimize parasitic absorption.
The inner tube is 6.0-inch outer diameter, 0.120-inch thick ASTM A269 tubing. Tolerances for ASTM A269 tubing of that diameter are +/- 0.030 inches for outer diameter and +/- 10% for wall thickness [1]. Thus, the inner tube is modeled with a 6.03-inch outer diameter (maximizing cavity volume and minimizing insulation) and 0.108-inch wall thickness. Similarly, the outer shell is 8.0-inch outer side length, 0.188-inch thick ASTM A554 tubing. Tolerances for ASTM A554 tubing of that size are +/- 0.060 inches for side length and +/- 10% for wall thickness [2].
Using a bounding tolerance of 0.1 inches for side length, the outer shell is modeled with a 7.9-inch outer side length and 0.169-inch wall thickness. Insulation is modeled as 1-inch thick, starting at the inner tube outer diameter and truncated as necessary to fit within the square outer shell. During HAC, the insulation may be lost and thus is not credited. No other changes are made to the ATR FFSC model during HAC as damage is limited to the not-modeled package ends.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                Rev. 16, May 2021 The fuel plates are assumed to be transported in a bag, for which the moderator density can vary independently from the rest of the ATR FFSC inner tube. The bag region is defined to be between the fuel plates and limited by the fuel plate widths. Up to 200 grams of polyethylene may be shipped in the cavity and assumed to possibly become homogenized with water inside the ATR FFSC cavity (either within or outside the bag region). Polyethylene is only credited when its presence is concluded to increase reactivity.
ATR FFSC model dimensions are summarized in Table 6.14-6.
6.14.3.1.2      ATR LEU Fuel Element Fuel meat volume is maximized by modeling the maximum allowable fuel meat arc length and height, while fuel plate volume (i.e. fuel meat plus cladding) is minimized by modeling the minimum fuel plate thickness. All other dimensions do not have explicitly specified tolerances and thus are modeled as nominal.
Plate-specific modeled dimensions are shown in Table 6.14-8. The modeled fuel meat height is 48.25 inches. The modeled NCT channel thicknesses either meet or exceed the nominal channel thicknesses. The modeled meat volume and U-235 mass for each fuel plate is shown in Table 6.14-9. The individual fuel meat U-235 masses are at least 2.3% greater than nominal masses, with the total U-235 mass 3.1% greater than nominal. This exceeds the +/- 2% tolerance on U-235 mass specified in Section 6.14.2.1. The modeled U-235 mass also exceeds the authorized maximum of 1,681 g (see Table 1.1-1).
Neither the LEU Balsa FHE nor existing ATR HEU FHE is modeled, but neoprene rub strips are accounted for based on the ATR HEU FHE design. The total volume of the neoprene rub strips is 62.65 in3. This is conservative since neoprene is not used in the ATR LEU FHE.
6.14.3.1.3      MURR LEU Fuel Element and DDE Fuel meat mass is increased beyond specified tolerances by first increasing the fuel meat density to a maximum possible density of 17.2 g/cm3 and then increasing the fuel meat thickness an additional 0.003 inches. This is done to bound possible adjustments to the fuel design. Fuel plate volume (i.e. fuel meat plus cladding) is minimized by modeling the minimum fuel plate thickness. All other dimensions do not have explicitly specified tolerances and thus are modeled as nominal.
Plate-specific modeled dimensions are shown in Table 6.14-10. The modeled fuel meat height is 24.2 inches. The modeled NCT channel thicknesses exceed nominal channel thicknesses. The modeled fuel meat U-235 mass for each fuel plate is shown in Table 6.14-11. The individual fuel meat U-235 masses are at least 5.0% greater than the respective nominal plus tolerance masses, with the total U-235 mass 9.4% greater than the total nominal plus tolerance mass. The modeled U-235 mass also exceeds the authorized maximum of 1,660 g (see Table 1.1-1).
Credit is taken for the MURR LEU FHE in restricting the movement of the fuel plates, but the aluminum material of the FHE is not modeled. Fuel plate movement is limited to within the confines of the FHE during both NCT and HAC. HAC damage to the FHE is modeled per Figure 2.12.3-4. The total volume of the FHE neoprene rub strips is 22.41 in3.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                              Rev. 16, May 2021 6.14.3.1.4      MIT LEU Fuel Element and DDE Fuel meat mass is increased beyond specified tolerances by first increasing the fuel meat density to a maximum possible density of 17.2 g/cm3 and then increasing the fuel meat thickness an additional 0.001 inches. This is done to bound possible adjustments to the fuel design. Fuel meat volume is maximized by modeling maximum fuel meat width and height. Fuel plate volume (i.e. fuel meat plus cladding) is minimized by modeling the minimum fuel plate thickness and width. Fuel plate thickness is reduced an additional 0.002 inches so that all channel thicknesses meet or exceed the maximum allowable average.
Plate-specific modeled dimensions are shown in Table 6.14-12. The modeled fuel meat height is 22.390 inches. The modeled NCT channel thicknesses exceed the nominal channel thicknesses.
The modeled fuel meat U-235 mass for each fuel plate is shown in Table 6.14-13. The individual fuel meat U-235 masses are at least 8.5% greater than the respective nominal plus tolerance masses, with the total U-235 mass 11.7% greater than the total nominal plus tolerance mass. The modeled U-235 mass also exceeds the authorized maximum of 1,070 g (see Table 1.1-1).
Credit is taken for the MIT FHE in restricting the movement of the fuel plates, but the aluminum material of the FHE is not modeled. Fuel plate movement is limited to within the confines of the FHE during both NCT and HAC. HAC damage to the FHE is modeled per Figure 2.12.3-3. The total volume of the FHE neoprene rub strips is 20.74 in3.
6.14.3.1.5      NBSR DDE Fuel meat mass is increased beyond the specified tolerances by increasing the fuel meat thickness an additional 0.001 inches. This is done to bound possible adjustments to the fuel design. Fuel meat volume is maximized by modeling maximum fuel meat width and height.
Fuel plate volume (i.e. fuel meat plus cladding) is minimized by modeling the minimum thickness and height.
Plate-specific modeled dimensions are shown in Table 6.14-7. The modeled U-235 mass is 14.16 g per fuel plate, resulting in 481 g U-235 per DDE. The total U-235 mass is 23% greater than the total nominal plus tolerance mass. The modeled U-235 mass also exceeds the authorized maximum of 460 g (see Table 1.1-1).
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                    Rev. 16, May 2021 Table 6.14 Key ATR FFSC Model Dimensions Item                    Nominal (in)      Modeled (in)
Inner round tube outer diameter        6.0                6.03 Inner round tube wall thickness        0.120              0.108 Insulation thickness                    1.0                1.0 Outer square tube outer side length    8.0                7.9 Outer square tube wall thickness        0.188              0.169 Table 6.14 Key NBSR LEU Model Dimensions Item                    Nominal (in)      Modeled (in)
Fuel Meat Thickness                    0.0085            0.0095 2.5030 Fuel Meat Flat-to-Flat Width                              2.5030 (prior to rolling)
Fuel Meat Height                        11.625            11.635 Fuel Plate Thickness                    0.050              0.048 2.7930 Fuel Plate Flat-to-Flat Width                              2.7930 (prior to rolling)
Fuel Plate Height                      13.000            12.010 Plate Curvature Radius                  5.500              5.500 Channel Thickness                      0.116              0.123 6-254
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                  Rev. 16, May 2021 Table 6.14 Key ATR LEU Model Dimensions Plate      Fuel Meat        Fuel Plate    Fuel Meat  Fuel Plate  NCT Channel
  #      Thickness (in)  Thickness (in)  Arc (deg)  Arc (deg)    Thickness (in) 1          0.008            0.079          34.7        40.0            -
2          0.013            0.049          35.2        39.1          0.080 3          0.013            0.049          35.5        39.3          0.079 4          0.016            0.049          35.9        39.5          0.078 5          0.016            0.049          36.2        39.8          0.079 6          0.016            0.049          36.5        39.9          0.079 7          0.016            0.049          36.8        40.1          0.079 8          0.016            0.049          37.1        40.3          0.079 9          0.016            0.049          37.3        40.4          0.079 10          0.016            0.049          37.5        40.5          0.079 11          0.016            0.049          37.8        40.7          0.079 12          0.016            0.049          38.0        40.8          0.079 13          0.016            0.049          38.2        40.9          0.079 14          0.016            0.049          38.4        41.0          0.078 15          0.016            0.049          38.5        41.1          0.079 16          0.013            0.049          38.7        41.2          0.079 17          0.008            0.049          38.9        41.3          0.079 18          0.008            0.049          39.0        41.4          0.079 19          0.008            0.099          39.2        42.2          0.078 Table 6.14 ATR LEU Fuel Meat Model Parameters Fuel Meat      U-235                  Fuel Meat      U-235 Plate #                                  Plate #
Volume (cm3)    Mass (g)              Volume (cm3)    Mass (g) 1          11.68        35.0          11        36.22        108.5 2          20.15        60.3          12        37.50        112.3 3          21.17        63.4          13        38.77        116.1 4          27.33        81.9          14        40.04        119.9 5          28.60        85.7          15        41.30        123.7 6          29.87        89.5          16        34.60        103.6 7          31.15        93.3          17        21.92          65.7 8          32.41        97.1          18        22.56          67.6 9          33.69        100.9          19        23.32          69.8 10          34.95        104.7        Total      567.24        1699.2 6-255
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                Rev. 16, May 2021 Table 6.14 Key MURR LEU Model Dimensions Plate    Fuel Meat        Fuel Plate    Fuel Meat Fuel Plate  NCT Channel
  #    Thickness (in)    Thickness (in)  Arc (deg) Arc (deg)    Thickness (in) 1        0.010              0.042        34.9      40.8              -
2        0.013              0.042        35.3      41.0            0.095 3        0.017              0.042        35.8      41.2            0.095 4        0.021              0.042        36.2      41.4            0.095 5        0.021              0.042        36.5      41.5            0.095 6        0.021              0.042        36.9      41.7            0.094 7        0.021              0.042        37.2      41.8            0.094 8        0.021              0.042        37.4      41.9            0.094 9        0.021              0.042        37.7      42.0            0.094 10        0.021              0.042        38.0      42.1            0.094 11        0.021              0.042        38.2      42.2            0.094 12        0.021              0.042        38.4      42.3            0.094 13        0.021              0.042        38.6      42.4            0.094 14        0.021              0.042        38.8      42.5            0.094 15        0.021              0.042        39.0      42.5            0.094 16        0.021              0.042        39.2      42.6            0.094 17        0.021              0.042        39.3      42.7            0.094 18        0.021              0.042        39.5      42.7            0.094 19        0.021              0.042        39.6      42.8            0.094 20        0.021              0.042        39.7      42.8            0.095 21        0.021              0.042        39.9      42.9            0.095 22        0.021              0.042        40.0      42.9            0.095 23        0.018              0.047        40.1      43.0            0.095 6-256
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                              Rev. 16, May 2021 Table 6.14 MURR LEU Fuel Meat Model Parameters U-235        Mass Above                  U-235    Mass Above Plate #                                    Plate #
Mass (g)  Max Allowable (g)              Mass (g) Max Allowable (g) 1      21.31            1.21              13      78.00        3.68 2      29.33            1.53              14      80.81        3.82 3      40.53            2.00              15      83.62        3.97 4      52.77            2.49              16      86.42        4.09 5      55.60            2.63              17      89.23        4.23 6      58.41            2.77              18      92.01        4.35 7      61.22            2.91              19      94.82        4.49 8      64.00            3.03              20      97.65        4.63 9      66.80            3.16              21    100.46        4.75 10      69.61            3.29              22    103.29        4.89 11      72.42            3.44              23      91.14        4.45 12      75.22            3.57            Total  1664.67      142.57 6-257
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                  Rev. 16, May 2021 Table 6.14 Key MIT LEU Model Dimensions Plate    Fuel Meat      Fuel Meat      Fuel Plate  Fuel Plate  NCT Channel
  #    Thickness (in)  Width (in)  Thickness (in)  Width (in)  Thickness (in) 1        0.014          2.177          0.044        2.521            -
2        0.018          2.177          0.044        2.521          0.080 3        0.018          2.177          0.044        2.521          0.079 4        0.026          2.177          0.044        2.521          0.080 5        0.026          2.177          0.044        2.521          0.080 6        0.026          2.177          0.044        2.521          0.079 7        0.026          2.177          0.044        2.521          0.080 8        0.026          2.177          0.044        2.521          0.079 9        0.026          2.177          0.044        2.521          0.080 10        0.026          2.177          0.044        2.521          0.080 11        0.026          2.177          0.044        2.521          0.079 12        0.026          2.177          0.044        2.521          0.080 13        0.026          2.177          0.044        2.521          0.079 14        0.026          2.177          0.044        2.521          0.080 15        0.026          2.177          0.044        2.521          0.080 16        0.026          2.177          0.044        2.521          0.079 17        0.018          2.177          0.044        2.521          0.080 18        0.018          2.177          0.044        2.521          0.079 19        0.014          2.177          0.044        2.521          0.080 Table 6.14 MIT LEU Fuel Meat Model Parameters U-235      Mass Above                    U-235        Mass Above Plate #                                    Plate #
Mass (g)  Max Allowable (g)              Mass (g)  Max Allowable (g) 1      34.92          3.85              11    64.85            5.09 2      44.90          4.27              12    64.85            5.09 3      44.90          4.27              13    64.85            5.09 4      64.85          5.09              14    64.85            5.09 5      64.85          5.09              15    64.85            5.09 6      64.85          5.09              16    64.85            5.09 7      64.85          5.09              17    44.90            4.27 8      64.85          5.09              18    44.90            4.27 9      64.85          5.09              19    34.92            3.85 10      64.85          5.09              Total  1092.45          115.75 6-258
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                      Rev. 16, May 2021 Figure 6.14 ATR LEU Single Package, NCT and HAC Figure 6.14 MURR LEU Single Package, NCT and HAC 6-259
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                        Rev. 16, May 2021 Figure 6.14 MIT LEU Single Package, NCT and HAC Figure 6.14 NBSR LEU Single Package, NCT and HAC 6-260
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                Rev. 16, May 2021 6.14.3.2 Material Properties In accordance with MCNP developer guidance, elemental compositions are converted to isotopic compositions using the mattool utility [3] included with MCNP6.2. An exception is made for materials with carbon, where elemental carbon is utilized as this cross-section is actively updated rather than carbon isotopes. For materials with oxygen, O-18 mass is redistributed into O-16 and O-17 masses since O-18 cross-sections are not available. Cross-sections available for use in MCNP6.2 are discussed in [4].
The fuel meat is constructed of U-10Mo fuel. The fuel is modeled such that uranium content is maximized (9% Mo, 91% wt.% U) and the enrichment is also maximized (19.95 wt.% U-235).
All other uranium isotopes are modeled at their respective limits with the remaining mass filled by U-238. The U-10Mo composition is shown in Table 6.14-14. The fuel meat density is 16.5 g/cm3 for the ATR LEU and 17.2 g/cm3 for all other fuel types. Fuel density is increased beyond nominal values, up to a manufacturing maximum of 17.2 g/cm3, to ensure fuel masses exceed respective tolerances and bound possible adjustments to fuel designs.
The fuel plate cladding is aluminum alloy 6061-O. The ATR FFSC structural tubing is type 304 stainless steel. The compositions for these materials are obtained from [5] and shown in Table 6.14-17 and Table 6.14-19 respectively.
Consistent with other analyses in this chapter, Durablanket S insulation is modeled as equal parts Al2O3 and SiO2 by weight. The density is 6 pounds per cubic foot (0.096 g/cm3). The Durablanket S composition is shown in Table 6.14-15.
Compositions for neoprene, polyethylene, and water are obtained from [5]. Water may be modeled as pure or homogenized with neoprene and/or polyethylene. Neoprene is conservatively modeled without chlorine, which is a neutron absorber. Neoprene density is adjusted to reflect the loss of chlorine. Materials are homogenized by (1) calculating the mass and volume of each sub-material, (2) calculating the isotopic masses for each sub-material using weight fractions and sub-material mass, and (3) summing isotopic masses into a homogenized material definition. The homogenized material density is calculated using the total mass and volume. The compositions for neoprene, polyethylene, and water are shown in Table 6.14-18, Table 6.14-20, and Table 6.14-16 respectively. For cases in which the homogenized material density is adjusted, the input water density is adjusted while the neoprene and/or polyethylene densities remain constant.
The composition of wood, for use in the air transport analysis, is taken from [5] and shown in Table 6.14-21. This material is discussed in Section 6.14.5.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                  Rev. 16, May 2021 Table 6.14 U-10Mo Composition Reference                    Modeled Element    ZAID    Wt. Fraction  ZAID    Wt. Fraction 42092    1.273450E-02 42094    8.131011E-03 42095    1.415610E-02 Mo      42000  9.00000E-02  42096    1.500685E-02 42097    8.690852E-03 42098    2.221755E-02 42100    9.063129E-03 U-232    92232  1.82000E-09  92232    1.820000E-09 U-234    92234  2.36600E-03  92234    2.366000E-03 U-235    92235  1.81545E-01  92235    1.815450E-01 U-236    92236  4.18600E-03  92236    4.186000E-03 U-238    92238  7.21903E-01  92238    7.219030E-01 Table 6.14 Durablanket S Composition (Density = 0.096 g/cm3)
Reference                    Modeled Element      ZAID  Wt. Fraction  ZAID    Wt. Fraction 8016  5.012956E-01 O        8000        50.2 8017  2.029452E-04 Al      13000        26.5      13027  2.647353E-01 14028  2.147528E-01 Si      14000        23.4      14029  1.129939E-02 14030  7.714015E-03 Table 6.14 Water Composition (Density = 1.0 g/cm3)
Reference                      Modeled Element    ZAID    Wt. Fraction  ZAID    Wt. Fraction 1001    1.118683E-01 H        1000    0.111894 1002    2.571290E-05 8016    8.877466E-01 O        8000    0.888106 8017    3.593967E-04 6-262
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                Rev. 16, May 2021 Table 6.14 Aluminum Alloy 6061-O Composition (Density = 2.7 g/cm3)
Reference                  Modeled Element    ZAID    Wt. Fraction ZAID  Wt. Fraction 12024  7.794921E-03 Mg      12000    0.010000    12025  1.028000E-03 12026  1.176980E-03 Al      13000    0.972000    13027  9.719903E-01 14028  5.511934E-03 Si    14000    0.006000    14029  2.900148E-04 14030  1.979911E-04 22046  6.969614E-05 22047  6.421984E-05 Ti    22000    0.000880    22048  6.498300E-04 22049  4.868279E-05 22050  4.756247E-05 24050  8.138607E-05 24052  1.632121E-03 Cr      24000    0.001950 24053  1.886331E-04 24054  4.784012E-05 Mn      25000    0.000880    25055  8.799912E-04 26054  2.309010E-04 26056  3.758735E-03 Fe      26000    0.004090 26057  8.835819E-05 26058  1.196495E-05 29063  1.883159E-03 Cu      29000    0.002750 29065  8.668135E-04 30064  6.887649E-04 30066  4.116611E-04 Zn      30000    0.001460    30067  6.127886E-05 30068  2.884319E-04 30070  9.848649E-06 Table 6.14 Neoprene Composition (No Chlorine, Density = 0.737 g/cm3)
Reference                  Modeled Element ZAID      Wt. Fraction ZAID  Wt. Fraction 1001  9.491352E-02 H      1000    0.056920 1002  2.181585E-05 C      6000    0.542646    6000  9.050647E-01 6-263
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                Rev. 16, May 2021 Table 6.14 Type 304 Stainless Steel Composition (Density = 8.00 g/cm3)
Reference                  Modeled Element    ZAID    Wt. Fraction ZAID  Wt. Fraction C      6000    0.000400    6000  3.999960E-04 14028  4.593278E-03 Si    14000    0.005000    14029  2.416790E-04 14030  1.649926E-04 P      15000    0.000230    15031  2.299977E-04 16032  1.420716E-04 16033  1.156799E-06 S      16000    0.000150 16034  6.753296E-06 16036  1.682534E-08 24050  7.929925E-03 24052  1.590272E-01 Cr      24000    0.190000 24053  1.837963E-02 24054  4.661345E-03 Mn      25000    0.010000    25055  9.999900E-03 26054  3.961618E-02 26056  6.448942E-01 Fe      26000    0.701730 26057  1.515980E-02 26058  2.052851E-03 28058  6.215726E-02 28060  2.476752E-02 Ni      28000    0.092500    28061  1.094596E-03 28062  3.547175E-03 28064  9.325298E-04 Table 6.14 Polyethylene Composition (Density = 0.930 g/cm3)
Reference                  Modeled Element ZAID      Wt. Fraction ZAID  Wt. Fraction 1001  1.436830E-01 H      1000    0.143716 1002  3.302550E-05 C      6000    0.856284    6000  8.562840E-01 6-264
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                Rev. 16, May 2021 Table 6.14 Wood Composition (Density = 0.64 g/cm3)
Reference                        Modeled Element    ZAID      Wt. Fraction    ZAID      Wt. Fraction 1001      5.962835E-02 H        1000      0.059642 1002      1.370556E-05 C        6000      0.497018      6000      4.970185E-01 7014      4.950631E-03 N        7000        0.00497 7015      1.937388E-05 8016      4.272625E-01 O        8000      0.427435 8017      1.729736E-04 12024      1.549647E-03 Mg      12000      0.001988      12025      2.043686E-04 12026      2.339861E-04 16032      4.707357E-03 16033      3.832904E-05 S      16000        0.00497 16034      2.237617E-04 16036      5.574856E-07 19039      1.847591E-03 K      19000      0.001988      19040      2.377463E-07 19041      1.401736E-04 20040      1.921638E-03 20042      1.346592E-05 20043      2.876710E-06 Ca      20000      0.001988 20044      4.548182E-05 20046      9.117818E-08 20048      4.447989E-06 6.14.3.3 Computer Codes and Cross-Section Libraries All criticality modeling is performed using MCNP6.2 [6]. Default .80c and .90c (H-1 only) continuous energy neutron cross-sections are utilized, which are based on ENDF/B-VII.1 and processed at 293.6 K. Thermal neutron S(, ) scattering data, based on ENDF/B-VII.0 and processed at 293.6 K, is applied to aluminum cladding (al27.22t, aluminum-27 metal), stainless steel (fe56.22t, iron-56 metal), and all pure or homogenized water (lwtr.20t, hydrogen in light water). All cases are run using 250 generations, with the first 50 generations skipped for calculation of keff and 10,000 neutrons per generation. These run parameters are sufficient to ensure proper convergence behavior for keff and Shannon entropy.
Calculation of USL is performed based on select cases using Whisper-1.1 [7] with Whisper default options. Cases are re-run in MCNP6.2 to calculate sensitivity coefficients using 600 generations, with the first 100 generations skipped and 10,000 neutrons per generation. The cases are analyzed in Whisper using 988 benchmarks (113 outliers identified by the developers are excluded from full set of 1,101 benchmarks).
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 6.14.3.4 Demonstration of Maximum Reactivity The U-235 mass is maximized within each fuel plate by maximizing fuel meat volume, uranium content, and enrichment without any offsetting reductions in fuel density or composition. Fuel plate cladding is minimized and structural components such as the fuel plate side plates, fuel plate end boxes, and FHE are not modeled except for moderating materials such as FHE neoprene. Volumes otherwise filled by these components can be filled by variable-density moderator or movement of the fuel plates. Moderating materials are allowed to become perfectly homogenized with water to maximize their effect.
No modeling of the ends of the ATR FFSC is performed. Modeling is axially limited to the maximum length of the fuel meat, reducing the amount of neutron absorbing materials. Axial water reflection begins at the ends of the fuel meat, maximizing the reflection of neutrons back into fuel.
For NCT, the fuel plates remain in a fixed spacing but may move freely as a group within the package cavity to maximize reactivity. For array cases, all fuel plates are shifted towards the center of the array. Water moderation is allowed within the package cavity because the package lid does not contain a seal. Since fuel elements may be transported in a thin bag, the level of moderation is allowed to vary independently between the fuel plates and outside the fuel plates but within the package cavity. Bag moderation is adjusted to have ideal moderation between fuel plates, while inner tube moderation is adjusted to either maximize reflection (single package) or balance leakage and moderation between packages (array cases). Moderating materials may become homogenized with water in either region (or not, if materials decrease reactivity).
For HAC, the side plates of the fuel elements are assumed to fail and the fuel plates are allowed to move independent of each other. Fuel plate spacing is adjusted to maximize reactivity. For array cases, all fuel plates are reoriented towards the center of the array. Water moderation is allowed within the package cavity or within the outer shell as damage may allow water into any region of the package. The insulation in the outer shell may be lost and thus is replaced by variable-density moderator. The bag holding the fuel plates is assumed to remain intact for altered fuel plate spacing. The level of moderation is allowed to vary independently between the fuel plates, outside the fuel plates but within the inner tube, and outside the inner tube but within the package outer shell. Similar to NCT, bag moderation is adjusted to have ideal moderation between fuel plates, while inner tube and outer shell moderation is adjusted to either maximize reflection (single package) or balance leakage and moderation between packages (array cases).
Moderating materials may become homogenized with water in the two regions within the package cavity.
Maximum reactivity for the air transport analysis is discussed in Section 6.14.5.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                    Rev. 16, May 2021 6.14.4 Package Criticality Calculations The package is considered to be acceptably subcritical if the computed ksafe, which is defined as keffective (keff) plus two standard deviations (), is less than or equal to the USL for all cases.
2 Results are reported using ksafe. The standard deviation for all cases is less than or equal to 0.00070. The maximum ksafe for each series of cases is listed in boldface.
6.14.4.1 NCT Single Package Evaluation 6.14.4.1.1          Configuration The NCT single package baseline configuration is the fuel plates centered in the cavity with full density pure water in both the fuel plate bag and inner tube. This configuration maximizes moderation between the fuel plates and reflection outside of the fuel plates. In the first series of cases, the water density inside the fuel plate bag and inside the inner tube are both lowered to confirm that maximum moderation and reflection are most conservative. In the second series of cases, the two regions of water are homogenized with neoprene and/or polyethylene (as applicable) to determine the most reactive locations for those materials.
6.14.4.1.2          Results NCT single package results are shown in Table 6.14-22 through Table 6.14-25. It can be seen that the fuel plates are under-moderated, and thus a decrease in fuel plate bag water density decreases ksafe. Similarly, a decrease in inner tube water density results in a decrease in reflection and therefore ksafe. With respect to neoprene and polyethylene, it is seen that the addition of polyethylene to either region increases ksafe while the addition of neoprene to either region decreases ksafe. The most reactive configuration is maximum density water in all regions and polyethylene in the fuel plate bag.
Table 6.14 ATR LEU NCT Single Package Results Bag Water            Inner Tube Water Neoprene                Polyethylene ksafe Density (g/cm3)          Density (g/cm ) 3 Location          Location Series 1: Water Density Adjustment 1                      1                    -                  -          0.42115 0.9                      1                    -                  -          0.40399 0.8                      1                    -                  -          0.38707 1                    0.9                  -                  -          0.40323 1                    0.8                  -                  -          0.38190 Series 2: Homogenized Material Location 1                      1                  Bag                -          0.40284 1                      1                    -                Bag          0.42385 1                      1                  Bag              Bag          0.40510 1                      1              Inner Tube              -          0.41764 1                      1                    -            Inner Tube        0.42207 1                      1              Inner Tube        Inner Tube        0.41820 6-267
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                        Rev. 16, May 2021 Table 6.14 MURR LEU NCT Single Package Results Bag Water      Inner Tube Water      Neoprene      Polyethylene ksafe Density (g/cm3)    Density (g/cm3)        Location        Location Series 1: Water Density Adjustment 1                    1                  -                -        0.50062 0.9                  1                  -                -        0.47599 0.8                  1                  -                -        0.45182 1                  0.9                -                -        0.48408 1                  0.8                -                -        0.46458 Series 2: Homogenized Material Location 1                    1                Bag                -        0.48870 1                    1                  -              Bag        0.50630 1                    1                Bag              Bag        0.49337 1                    1            Inner Tube            -        0.49801 1                    1                  -          Inner Tube    0.50184 1                    1            Inner Tube      Inner Tube    0.49978 Table 6.14 MIT LEU NCT Single Package Results Bag Water      Inner Tube Water Neoprene            Polyethylene ksafe Density (g/cm3)    Density (g/cm )3 Location        Location Series 1: Water Density Adjustment 1                    1                -                -          0.42227 0.9                  1                -                -          0.40764 0.8                  1                -                -          0.39258 1                  0.9                -                -          0.40197 1                  0.8                -                -          0.38075 Series 2: Homogenized Material Location 1                    1                Bag              -          0.40892 1                    1                -              Bag        0.42693 1                    1                Bag              Bag        0.41531 1                    1            Inner Tube            -          0.42058 1                    1                -          Inner Tube      0.42349 1                    1            Inner Tube      Inner Tube      0.42153 6-268
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 Table 6.14 NBSR LEU NCT Single Package Results Bag Water          Inner Tube Water        Polyethylene ksafe Density (g/cm3)        Density (g/cm3)            Location Series 1: Water Density Adjustment 1                      1                      -            0.38796 0.9                      1                      -            0.36870 0.8                      1                      -            0.34843 1                      0.9                    -            0.37027 1                      0.8                    -            0.35170 Series 2: Homogenized Material Location 1                      1                    Bag            0.39137 1                      1                Inner Tube        0.38896 6.14.4.2 NCT Package Array Evaluation 6.14.4.2.1      Configuration The NCT package array baseline configuration is an array of packages, with fuel plates in each package shifted towards the center of the array. Full density pure water is in both the fuel plate bag and inner tube. In the first series of cases, the water density in the inner tube is first lowered to find the optimal level of moderation between packages. Then, the water density in the fuel plate bag is lowered to confirm that the individual fuel elements are under-moderated and thus maximum density moderator between the fuel plates is most reactive. In the second series of cases, the two regions of water are homogenized with neoprene and/or polyethylene (as applicable) to determine the most reactive locations for those materials. Inner tube water density is adjusted after homogenization to verify the optimal density is being used.
The package array configurations are shown in Figure 6.14-5 through Figure 6.14-8. The size of the NCT array is 7x7x1 for ATR LEU, 8x8x1 for MURR LEU, and 9x9x1 for MIT LEU and NBSR LEU.
6.14.4.2.2      Results NCT package array results are shown in Table 6.14-26 through Table 6.14-29. It can be seen that a significantly reduced inner tube water density, optimizing moderation and leakage between packages, results in the highest ksafe. Like the NCT single package configuration, a reduction in fuel plate bag water density decreases ksafe since the individual fuel elements are under-moderated. With respect to neoprene and polyethylene, ksafe is maximized when these materials are homogenized with water within the package (though the effect is relatively small).
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                        Rev. 16, May 2021 Figure 6.14 ATR LEU Package Array, NCT and HAC Figure 6.14 MURR LEU Package Array, NCT and HAC 6-270
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                        Rev. 16, May 2021 Figure 6.14 MIT LEU Package Array, NCT and HAC Figure 6.14 NBSR LEU Package Array, NCT and HAC 6-271
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                        Rev. 16, May 2021 Table 6.14 ATR LEU NCT Package Array Results Bag Water      Inner Tube Water Neoprene            Polyethylene ksafe Density (g/cm3)    Density (g/cm3)      Location        Location Series 1: Water Density Adjustment 1                  1                  -                -          0.73682 1                  0.7                -                -          0.76281 1                  0.6                -                -          0.77094 1                  0.5                -                -          0.77482 1                  0.4                -                -          0.77288 1                  0.3                -                -          0.76351 0.9                0.5                -                -          0.75601 0.8                0.5                -                -          0.73860 Series 2: Homogenized Material Location 1                  0.5              Bag                -          0.75463 1                  0.5                -              Bag          0.77684 1                  0.5              Bag              Bag          0.75681 1                  0.3            Inner Tube            -          0.76843 1                  0.4            Inner Tube            -          0.77572 1                  0.5            Inner Tube            -          0.77569 1                  0.6            Inner Tube            -          0.77137 1                  0.3                -          Inner Tube      0.76572 1                  0.4                -          Inner Tube      0.77509 1                  0.5                -          Inner Tube      0.77481 1                  0.6                -          Inner Tube      0.77063 1                  0.3            Inner Tube      Inner Tube      0.77090 1                  0.4            Inner Tube      Inner Tube      0.77525 1                  0.5            Inner Tube      Inner Tube      0.77530 1                  0.6            Inner Tube      Inner Tube      0.77091 1                  0.3            Inner Tube          Bag          0.77218 1                  0.4          Inner Tube          Bag          0.77809 1                  0.5            Inner Tube          Bag          0.77724 6-272
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                        Rev. 16, May 2021 Table 6.14 MURR LEU NCT Package Array Results Bag Water      Inner Tube Water Neoprene            Polyethylene ksafe Density (g/cm3)    Density (g/cm3)      Location        Location Series 1: Water Density Adjustment 1                    1                -                -          0.83341 1                  0.8                -                -          0.84791 1                  0.7                -                -          0.85406 1                  0.6                -                -          0.85798 1                  0.5                -                -          0.85697 1                  0.4                -                -          0.85219 0.9                  0.6                -                -          0.83604 0.8                  0.6                -                -          0.81421 Series 2: Homogenized Material Location 1                  0.6              Bag              -          0.84759 1                  0.6                -              Bag          0.85992 1                  0.6              Bag            Bag          0.85000 1                  0.4          Inner Tube            -          0.85465 1                  0.5          Inner Tube            -          0.85904 1                  0.6          Inner Tube            -          0.85672 1                  0.7          Inner Tube            -          0.85569 1                  0.4                -          Inner Tube      0.85380 1                  0.5                -          Inner Tube      0.85809 1                  0.6                -          Inner Tube      0.85663 1                  0.7                -          Inner Tube      0.85332 1                  0.4          Inner Tube      Inner Tube      0.85631 1                  0.5          Inner Tube      Inner Tube      0.85806 1                  0.6          Inner Tube      Inner Tube      0.85791 1                  0.7          Inner Tube      Inner Tube      0.85291 1                  0.4          Inner Tube          Bag          0.85749 1                  0.5          Inner Tube          Bag          0.86146 1                  0.6          Inner Tube          Bag          0.86127 6-273
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                      Rev. 16, May 2021 Table 6.14 MIT LEU NCT Package Array Results Bag Water      Inner Tube Water      Neoprene      Polyethylene ksafe Density (g/cm3)    Density (g/cm3)        Location        Location Series 1: Water Density Adjustment 1                  1                  -              -        0.65761 1                  0.8                  -              -        0.67282 1                  0.7                  -              -        0.67735 1                  0.6                  -              -        0.68189 1                  0.5                  -              -        0.68170 1                  0.4                  -              -        0.67448 0.9                0.6                  -              -        0.66814 0.8                0.6                  -              -        0.65564 Series 2: Homogenized Material Location 1                  0.6                Bag              -        0.66801 1                  0.6                  -            Bag        0.68549 1                  0.6                Bag            Bag        0.67260 1                  0.4            Inner Tube          -        0.67691 1                  0.5            Inner Tube          -        0.68206 1                  0.6            Inner Tube          -        0.68140 1                  0.7            Inner Tube          -        0.67777 1                  0.4                  -          Inner Tube    0.67604 1                  0.5                  -          Inner Tube    0.68150 1                  0.6                  -          Inner Tube    0.68124 1                  0.7                  -          Inner Tube    0.67652 1                  0.4            Inner Tube      Inner Tube    0.67843 1                  0.5            Inner Tube      Inner Tube    0.68258 1                  0.6            Inner Tube      Inner Tube    0.68096 1                  0.7            Inner Tube      Inner Tube    0.67788 1                  0.4            Inner Tube        Bag        0.68211 1                  0.5            Inner Tube        Bag        0.68563 1                  0.6            Inner Tube          Bag        0.68613 1                  0.7            Inner Tube        Bag        0.68210 6-274
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                  Rev. 16, May 2021 Table 6.14 NBSR LEU NCT Package Array Results Bag Water        Inner Tube Water        Polyethylene ksafe Density (g/cm3)      Density (g/cm3)        Location Series 1: Water Density Adjustment 1                    1                    -    0.63207 1                  0.8                  -    0.64205 1                  0.7                  -    0.64650 1                  0.6                  -    0.65053 1                  0.5                  -    0.65048 1                  0.4                  -    0.64692 0.9                  0.6                  -    0.63072 0.8                  0.6                  -    0.61406 Series 2: Homogenized Material Location 1                  0.6                Bag    0.65257 1                  0.7              Inner Tube 0.64567 1                  0.6              Inner Tube 0.64911 1                  0.5              Inner Tube 0.65005 1                  0.4              Inner Tube 0.64878 1                  0.3              Inner Tube 0.64213 6-275
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                    Rev. 16, May 2021 6.14.4.3 HAC Single Package Evaluation 6.14.4.3.1      Configuration The HAC single package baseline configuration is the fuel plates centered in the cavity with full density pure water in the fuel plate bag, inner tube, and outer shell and normal spacing between plates. In the first series of cases, the spacing between fuel plates is adjusted to determine the optimal spacing. In the second series of cases, the water density in each region is lowered to confirm that maximum moderation and reflection are most conservative. In the third series of cases, the two regions of water in the package cavity are homogenized with neoprene and/or polyethylene (as applicable) to determine the most reactive locations for those materials.
For MIT LEU, the fuel plates are horizontally aligned (rather than diagonally offset) and become only slightly offset as necessary to stay within the FHE envelope (see Figure 6.14-3).
For NBSR LEU, additional series of cases are performed to evaluate changes to the axial spacing between the two sets of 17 fuel plates. First, the axial spacing between the two sets of plates in reduced to determine the optimal spacing. Second, a single case is performed in which the two sets are allowed to perfectly overlap, resulting in a compressed assembly with smaller channel gaps.
6.14.4.3.2      Results HAC single package results are shown in Table 6.14-30 through Table 6.14-37. As expected based on the NCT single package results, the fuel plates are under-moderated and ksafe increases as the distance between fuel plates increases. For NBSR LEU, ksafe is maximized with minimum axial spacing (without overlapping) and maximum transverse spacing (i.e. maximum channel thicknesses). For water density, the most conservative configuration is maximum density water in all locations. With respect to neoprene and polyethylene, ksafe is maximized when only polyethylene is credited.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                                              Rev. 16, May 2021 Table 6.14 ATR LEU HAC Single Package Results (No Additional Materials)
Bag Water        Inner Tube Water Outer Shell Water              Additional Fuel ksafe Density (g/cm3)      Density (g/cm3)        Density (g/cm3)      Plate Spacing (cm)
Series 1: Fuel Plate Spacing 1                      1                      1                      0.00            0.45162 1                      1                      1                    -0.10            0.39185 1                      1                      1                      0.10            0.50095 1                      1                      1                      0.20            0.54128 1                      1                      1                      0.30            0.57217 1                      1                      1                      0.36            0.58803 Series 2: Water Density Adjustment 1                      1                      1                      0.36            0.58803 0.9                    1                      1                      0.36            0.55434 0.8                    1                      1                      0.36            0.52135 1                    0.9                    1                      0.36            0.57928 1                    0.8                    1                      0.36            0.56997 1                      1                    0.9                    0.36            0.58388 1                      1                    0.8                    0.36            0.57894 Table 6.14 ATR LEU HAC Single Package Results (Constant Spacing)
Bag Water        Inner Tube Water        Outer Shell Water        Neoprene      Polyethylene ksafe Density (g/cm3)      Density (g/cm3)          Density (g/cm3)        Location          Location Series 3: Homogenized Material Location (0.36 cm spacing) 1                    1                      1                  Bag                -          0.57518 1                    1                      1                    -              Bag          0.58819 1                    1                      1                  Bag              Bag          0.57690 1                    1                      1              Inner Tube            -          0.57660 1                    1                      1                    -          Inner Tube        0.58853 1                    1                      1              Inner Tube        Inner Tube      0.57753 6-277
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                                          Rev. 16, May 2021 Table 6.14 MURR LEU HAC Single Package Results (No Additional Materials)
Bag Water        Inner Tube Water Outer Shell Water              Additional Fuel ksafe Density (g/cm3)      Density (g/cm3)        Density (g/cm3)      Plate Spacing (cm)
Series 1: Fuel Plate Spacing 1                    1                      1                    0.000          0.53856 1                    1                      1                  -0.100          0.47550 1                    1                      1                    0.025          0.55215 1                    1                      1                    0.050          0.56541 1                    1                      1                    0.075          0.57805 1                    1                      1                    0.081          0.58197 Series 2: Water Density Adjustment 1                    1                      1                    0.081          0.58197 0.9                  1                      1                    0.081          0.55333 0.8                  1                      1                    0.081          0.52335 1                  0.9                      1                    0.081          0.56998 1                  0.8                      1                    0.081          0.55795 1                    1                      0.9                  0.081          0.57875 1                    1                      0.8                  0.081          0.57471 Table 6.14 MURR LEU HAC Single Package Results (Constant Spacing)
Bag Water      Inner Tube Water Outer Shell Water Neoprene                  Polyethylene ksafe Density (g/cm3)    Density (g/cm3)        Density (g/cm3)      Location        Location Series 3: Homogenized Material Location (0.081 cm spacing) 1                  1                      1                Bag              -        0.57057 1                  1                      1                  -            Bag        0.58572 1                  1                      1                Bag            Bag        0.57448 1                  1                      1            Inner Tube          -        0.57999 1                  1                      1                  -          Inner Tube    0.58157 1                  1                      1            Inner Tube      Inner Tube    0.58069 6-278
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                                              Rev. 16, May 2021 Table 6.14 MIT LEU HAC Single Package Results (No Additional Materials)
Bag Water        Inner Tube Water Outer Shell Water                Additional Fuel ksafe Density (g/cm3)      Density (g/cm3)          Density (g/cm3)      Plate Spacing (cm)
Series 1: Fuel Plate Spacing 1                      1                      1                    0.00          0.44515 1                      1                      1                    -0.05          0.41808 1                      1                      1                    0.05          0.46905 1                      1                      1                    0.10          0.49110 1                      1                      1                    0.15          0.51203 1                      1                      1                    0.16          0.51582 Series 2: Water Density Adjustment 1                      1                      1                    0.16          0.51582 0.9                    1                      1                    0.16          0.49164 0.8                    1                      1                    0.16          0.46852 1                    0.9                      1                    0.16          0.50219 1                    0.8                      1                    0.16          0.48752 1                      1                      0.9                    0.16          0.51320 1                      1                      0.8                    0.16          0.51086 Table 6.14 MIT LEU HAC Single Package Results (Constant Spacing)
Bag Water      Inner Tube Water Outer Shell Water Neoprene                    Polyethylene ksafe Density (g/cm3)    Density (g/cm3)          Density (g/cm3)      Location        Location Series 3: Homogenized Material Location (0.16 cm spacing) 1                    1                      1                Bag                -          0.50429 1                    1                      1                  -              Bag          0.52045 1                    1                      1                Bag              Bag          0.50864 1                    1                      1              Inner Tube            -          0.51554 1                    1                      1                  -          Inner Tube      0.51619 1                    1                      1              Inner Tube      Inner Tube      0.51461 6-279
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                      Rev. 16, May 2021 Table 6.14 NBSR LEU HAC Single Package Results (No Additional Materials)
Additional Fuel            Axial Gap ksafe Plate Spacing (cm) Reduction (cm)
Series 1: Fuel Plate Spacing 0.00                      0        0.41491
                                      -0.10                      0        0.37573 0.10                      0        0.44508 0.20                      0        0.46574 0.30                      0        0.48013 0.35                      0        0.48391 Series 1.a: Axial Gap Reduction 0.35                      0.0      0.48391 0.35                    -1.0      0.48216 0.35                      1.0      0.48720 0.35                      2.0      0.49136 0.35                      3.0      0.49496 0.35                      4.0      0.50060 0.35                      5.0      0.50674 Series 1.b: Overlapping Plate Sets
                                          -                        -        0.49491 6-280
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                                Rev. 16, May 2021 Table 6.14 NBSR LEU HAC Single Package Results (Constant Spacing)
Bag Water      Inner Tube Water Outer Shell Water Polyethylene ksafe Density (g/cm3)  Density (g/cm3)        Density (g/cm3)      Location Series 4: Water Density Adjustment (Maximum Transverse Spacing, Minimum Axial Gap) 1                  1                      1                -    0.50674 0.9                1                      1                -    0.47818 0.8                1                      1                -    0.45045 1                0.9                    1                -    0.49769 1                0.8                    1                -    0.48948 1                  1                    0.9                -    0.50263 1                  1                    0.8                -    0.49941 Series 5: Homogenized Material Location (Maximum Transverse Spacing, Minimum Axial Gap) 1                  1                      1              Bag    0.50968 1                  1                      1            Inner Tube 0.50820 6-281
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                    Rev. 16, May 2021 6.14.4.4 HAC Package Array Evaluation 6.14.4.4.1      Configuration The HAC package array baseline configuration is an array of packages, with fuel plates in each package angled towards the center of the array and full density pure water in the fuel plate bag, inner tube, and outer shell regions. Based on the HAC single package results, fuel plates are maximally spaced. For NBSR LEU, fuel plates are maximally spaced with the minimum axial gap. In the first series of cases, the water densities in the outer shell and inner tube are lowered to find the optimal level of moderation between packages. Then, the water density in the fuel plate bag is lowered to confirm that the individual fuel elements are under-moderated and thus maximum density moderator is most reactive. In the second series of cases, the two regions of water in the package cavity are homogenized with neoprene and/or polyethylene (as applicable) to determine the most reactive locations for those materials. Inner tube water density is adjusted after homogenization to verify the optimal density is being used.
The size of the HAC array is 4x4x1 for ATR LEU and 5x5x1 for MURR LEU, MIT LEU, and NBSR LEU.
6.14.4.4.2      Results HAC package array results are shown in Table 6.14-38 through Table 6.14-45. A significant reduction in inner tube and outer shell water density is required to optimize moderation. A significantly higher ksafe results from first reducing outer shell water density to zero and then reducing inner tube water density rather than the reverse. Like the HAC single package configuration, a reduction in fuel plate bag water density decreases ksafe showing that the individual fuel elements are under-moderated. With respect to neoprene and polyethylene, ksafe is maximized when these materials are homogenized with water within the package (though the effect is relatively small).
6-282
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                        Rev. 16, May 2021 Table 6.14 ATR LEU HAC Package Array Results (No Additional Materials)
Bag Water        Inner Tube Water Outer Shell Water ksafe Density (g/cm3) Density (g/cm3)          Density (g/cm3)
Series 1: Water Density Adjustment (0.36 cm spacing) 1                    1                      1            0.70983 1                  0.5                    1            0.69955 1                    0                      1            0.69555 1                    0                    0.6          0.75451 1                    0                    0.3          0.80200 1                    0                    0.2          0.81176 1                    0                    0.1          0.81225 1                    0                      0            0.79948 1                    1                      0            0.87577 1                  0.9                    0            0.87718 1                  0.8                    0            0.87710 1                  0.7                    0            0.87752 1                  0.6                    0            0.87503 1                  0.5                    0            0.87128 0.9                  0.7                    0            0.84911 0.8                  0.7                    0            0.81838 6-283
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                        Rev. 16, May 2021 Table 6.14 ATR LEU HAC Package Array Results (Void in Outer Shell)
Bag Water        Inner Tube Water Neoprene            Polyethylene ksafe Density (g/cm3)    Density (g/cm3)      Location        Location Series 2: Homogenized Material Location (0.36 cm spacing) 1                  0.7              Bag                -          0.86677 1                  0.7                -              Bag          0.87765 1                  0.7              Bag              Bag          0.86865 1                  0.6            Inner Tube            -          0.87575 1                  0.7            Inner Tube            -          0.87703 1                  0.8            Inner Tube            -          0.87789 1                  0.9            Inner Tube            -          0.87879 1                  1            Inner Tube            -          0.87693 1                  0.6                -          Inner Tube      0.87578 1                  0.7                -          Inner Tube      0.87737 1                  0.8                -          Inner Tube      0.87706 1                  0.9                -          Inner Tube      0.87699 1                  1                  -          Inner Tube      0.87510 1                  0.6            Inner Tube      Inner Tube      0.87656 1                  0.7            Inner Tube      Inner Tube      0.87889 1                  0.8            Inner Tube      Inner Tube      0.87978 1                  0.9            Inner Tube      Inner Tube      0.87818 1                  1            Inner Tube      Inner Tube      0.87575 1                  0.8            Inner Tube          Bag          0.87877 1                  0.9          Inner Tube          Bag          0.87986 1                  1            Inner Tube          Bag          0.87661 6-284
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                        Rev. 16, May 2021 Table 6.14 MURR LEU HAC Package Array Results (No Additional Materials)
Bag Water        Inner Tube Water Outer Shell Water ksafe Density (g/cm3)      Density (g/cm3)      Density (g/cm3)
Series 1: Water Density Adjustment (0.081 cm spacing) 1                    1                      1            0.69387 1                    0.5                    1            0.68303 1                    0                      1            0.70321 1                    0                    0.5          0.77612 1                    0                    0.3          0.79783 1                    0                    0.2          0.79887 1                    0                    0.1          0.78621 1                    0                      0            0.75823 1                    1                      0            0.85815 1                    0.9                    0            0.86123 1                    0.8                    0            0.86222 1                    0.7                    0            0.86123 1                    0.6                    0            0.85913 0.9                  0.8                    0            0.83494 0.8                  0.8                    0            0.81087 6-285
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                      Rev. 16, May 2021 Table 6.14 MURR LEU HAC Package Array Results (Void in Outer Shell)
Bag Water        Inner Tube Water Neoprene          Polyethylene ksafe Density (g/cm3)    Density (g/cm3)    Location          Location Series 2: Homogenized Material Location (0.081 cm spacing) 1                  0.8              Bag              -          0.85277 1                  0.8                -              Bag          0.86412 1                  0.8              Bag            Bag          0.85625 1                  0.6          Inner Tube            -          0.85909 1                  0.7          Inner Tube            -          0.86102 1                  0.8          Inner Tube            -          0.86181 1                  0.9          Inner Tube            -          0.86113 1                  1            Inner Tube            -          0.86038 1                  0.6                -          Inner Tube      0.86038 1                  0.7                -          Inner Tube      0.86159 1                  0.8                -          Inner Tube      0.86171 1                  0.9                -          Inner Tube      0.85956 1                  1                -          Inner Tube      0.85777 1                  0.6          Inner Tube      Inner Tube      0.86025 1                  0.7          Inner Tube      Inner Tube      0.86322 1                  0.8          Inner Tube      Inner Tube      0.86236 1                  0.9          Inner Tube      Inner Tube      0.86153 1                  1            Inner Tube      Inner Tube      0.85890 1                  0.7          Inner Tube          Bag          0.86481 1                  0.8          Inner Tube          Bag          0.86625 1                  0.9          Inner Tube          Bag          0.86565 6-286
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                        Rev. 16, May 2021 Table 6.14 MIT LEU HAC Package Array Results (No Additional Materials)
Bag Water        Inner Tube Water Outer Shell Water ksafe Density (g/cm3)      Density (g/cm3)      Density (g/cm3)
Series 1: Water Density Adjustment (0.16 cm spacing) 1                    1                      1          0.59418 1                  0.5                    1          0.56734 1                    0                      1          0.57223 1                    0                    0.5          0.62684 1                    0                    0.4          0.63504 1                    0                    0.3          0.64180 1                    0                    0.2          0.63851 1                    0                    0.1          0.62576 1                    1                      0          0.71703 1                  0.9                    0          0.72087 1                  0.8                    0          0.72086 1                  0.7                    0          0.72276 1                  0.6                    0          0.71887 1                  0.5                    0          0.71277 0.9                  0.7                    0          0.69897 0.8                  0.7                    0          0.67818 6-287
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                        Rev. 16, May 2021 Table 6.14 MIT LEU HAC Package Array Results (Void in Outer Shell)
Bag Water        Inner Tube Water Neoprene            Polyethylene ksafe Density (g/cm3)    Density (g/cm3)      Location        Location Series 2: Homogenized Material Location (0.16 cm spacing) 1                  0.7              Bag                -          0.71042 1                  0.7                -              Bag          0.72449 1                  0.7              Bag              Bag          0.71468 1                  0.5            Inner Tube            -          0.71423 1                  0.6            Inner Tube            -          0.71903 1                  0.7            Inner Tube            -          0.72189 1                  0.8            Inner Tube            -          0.72216 1                  0.9            Inner Tube            -          0.72228 1                  1            Inner Tube            -          0.71873 1                  0.5                -          Inner Tube      0.71410 1                  0.6                -          Inner Tube      0.71825 1                  0.7                -          Inner Tube      0.72301 1                  0.8                -          Inner Tube      0.72222 1                  0.9                -          Inner Tube      0.72156 1                  1                  -          Inner Tube      0.71655 1                  0.5            Inner Tube      Inner Tube      0.71487 1                  0.6            Inner Tube      Inner Tube      0.71970 1                  0.7            Inner Tube      Inner Tube      0.72231 1                  0.8            Inner Tube      Inner Tube      0.72289 1                  0.9            Inner Tube      Inner Tube      0.72117 1                  1            Inner Tube      Inner Tube      0.71900 1                  0.7            Inner Tube          Bag          0.72656 1                  0.8          Inner Tube          Bag          0.72735 1                  0.9            Inner Tube          Bag          0.72551 1                  1            Inner Tube          Bag          0.72280 6-288
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                      Rev. 16, May 2021 Table 6.14 NBSR LEU HAC Package Array Results (No Additional Materials)
Bag Water        Inner Tube Water Outer Shell Water ksafe Density (g/cm3)      Density (g/cm3)        Density (g/cm3)
Series 1: Water Density Adjustment (Maximum Transverse Spacing, Minimum Axial Gap) 1                      1                      1        0.61401 1                    0.5                    1        0.60045 1                      0                      0        0.66893 1                      0                    0.5        0.64935 1                      0                    0.4        0.66184 1                      0                    0.3        0.67187 1                      0                    0.2        0.67958 1                      0                    0.1        0.67985 1                      0                      0        0.66893 1                      1                    0.1        0.64668 1                      1                      0        0.74301 1                    0.9                    0        0.74621 1                    0.8                    0        0.74596 1                    0.7                    0        0.74612 1                    0.6                    0        0.74448 1                    0.5                    0        0.74017 0.9                    0.9                    0        0.72321 0.8                    0.9                    0        0.69974 Table 6.14 NBSR LEU HAC Package Array Results (Void in Outer Shell)
Bag Water        Inner Tube Water Polyethylene 3                                            ksafe Density (g/cm )      Density (g/cm3)        Location Series 2: Homogenized Material Location (Maximum Transverse Spacing, Minimum Axial Gap) 1                    0.9                Bag      0.74795 1                    1              Inner Tube  0.74325 1                    0.9            Inner Tube  0.74599 1                    0.8            Inner Tube    0.74897 1                    0.7            Inner Tube  0.74840 1                    0.6            Inner Tube  0.74569 6-289
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                Rev. 16, May 2021 6.14.5 Fissile Material Packages for Air Transport To demonstrate compliance with the requirements for air transport of fissile material in 10 CFR 71.55(f), a separate analysis is performed in which any structural integrity of the package is not credited. The fissile material from a single package is modeled as a sphere that is moderated solely by package materials and may be reflected by other package materials and water. This methodology is consistent with the air transport analysis in Section 6.7. This analysis uses conservative material quantities to bound ATR, MURR, MIT, and NBSR LEU fuel elements and/or DDEs.
6.14.5.1 Configuration The air transport configuration is a sphere of fissile material surrounded by 20 cm of water. For air transport analysis, the fissile sphere may be moderated by package materials but it is assumed that there is no water inleakage [8]. To optimally moderate the fissile sphere, a bounding quantity of uranium fuel is mixed with increasing amounts of moderating material that may be present in the ATR FFSC. Moderating materials are neoprene, polyethylene, and wood. Wood represents either structural materials (such as the ATR LEU Balsa FHE) or cellulosic materials (such as cardboard dunnage). The wood density used (see Table 6.14-21) is for relatively dense wood and is higher than that for balsa wood or cellulosic materials. This is conservative as it results in a smaller fissile sphere volume for the same wood mass, minimizing neutron leakage.
Only pure water is used as a reflector as there are no package structural materials which would function as a superior reflector.
Two series of cases are performed. In the first series, increasing amounts of moderating material are added to the fissile sphere to optimize moderation. In the second series, the fissile sphere is reconfigured to consist of two regions: a central fuel-moderator sphere and an outer fuel-only layer. The total mass of fuel is kept constant in the second series. The single-region configuration is shown in Figure 6.14-9, with the fuel-moderator region in red and the surrounding 20 cm of water in blue.
To bound any enrichment of U-235, the uranium fuel is modeled as U-235 only. The density of U-235 is set to 18.76 g/cm3, which is calculated based on the density of pure U-233 (18.60 g/cm3 per [9]) multiplied by the ratio of mass numbers. Material limits are 2000 g U-235, 1500 g neoprene, 200 g polyethylene, and 4000 g wood material. The neoprene mass limit represents real-world neoprene mass (i.e. neoprene without chlorine removed) while modeling uses conservative, reduced mass neoprene with chlorine removed.
6.14.5.2 Results Air transport results are shown in Table 6.14-46. It can be seen that the single region fissile sphere is under-moderated and thus ksafe increases as moderating materials are added to the sphere. For the two-region fissile sphere, moving U-235 to the outer region increases ksafe initially before eventually peaking at an even split of U-235.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                  Rev. 16, May 2021 Figure 6.14 Air Transport Model, Single Region Table 6.14 Air Transport Results Inner Region    Outer Region Neoprene Polyethylene            Wood ksafe U-235 Mass (g)  U-235 Mass (g) Mass (g)            Mass (g)    Mass (g)
Series 1: Increasing Moderating Material 2000              -                0            0          0      0.52101 2000              -                0            100        0      0.52978 2000              -                0            200        0      0.56477 2000              -              500          200        0      0.57453 2000              -              1000          200        0      0.60509 2000              -              1500          200        0      0.63379 2000              -              1500          200      1000      0.63832 2000              -              1500          200      2000      0.65819 2000              -              1500          200      3000      0.67665 2000              -              1500          200      4000      0.69452 Series 2: Two Regions U-235 1900              100            1500          200      4000      0.69936 1500              500            1500          200      4000      0.70584 1200              800            1500          200      4000      0.70846 1100              900            1500          200      4000      0.70928 1000            1000            1500          200      4000      0.71171 900            1100            1500          200      4000      0.70911 800            1200            1500          200      4000      0.70790 500            1500            1500          200      4000      0.70398 100            1900            1500          200      4000      0.67082 6-291
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 6.14.6 Benchmark Evaluations Whisper-1.1 [7], a statistical analysis package distributed with MCNP6.2, is used for USL calculation. Whisper uses sensitivity profile data, as computed by MCNP, along with nuclear data covariance files to calculate USL for an input application. Sensitivity/uncertainty (S/U) techniques are used to guide the selection of benchmarks similar to the application.
Per [7], MCNP6-Whisper provides repeatable, quantitative, physics-based information to NCS analysts for determining USLs, replacing much of what used to be carried out based solely on expert judgment. Whisper-1.1 is maintained under the same SQA procedures as MCNP [10].
A detailed explanation of Whispers underlying theory is documented in [11].
The cases which result in the maximum ksafe for each fuel type are selected for evaluation with Whisper (worst-case HAC package array for all types). Additionally, since the air transport configuration is significantly different than ATR FFSC configurations, the air transport case which results in a maximum ksafe (two regions, maximum moderating materials, even split of U-235 between regions) is also evaluated. The cases are re-run in MCNP to generate sensitivity profiles, which are then input into Whisper for comparison to benchmark sensitivity profiles.
For MURR LEU, both the worst-case NCT package array and HAC package array cases are analyzed with Whisper since these cases have comparable ksafe values. It is determined that the NCT package array case results in a lower USL and thus instead serves as the basis for the MURR LEU USL.
6.14.6.1 Applicability of Benchmark Experiments Whisper-1.1 comes pre-packaged with 1,101 benchmarks, each with an MCNP input file and MCNP-generated sensitivity profile. The benchmarks based on experiments documented in the International Handbook of Evaluated Criticality Safety Benchmark Experiments [12] and are identified by experiment title and case number. The benchmarks inputs have been evaluated using the same ENDF/B-VII.1, 293.6 K continuous energy neutron cross-sections as were used for ATR FFSC criticality modeling. 113 benchmarks have been identified by the Whisper developers as statistical outliers and are excluded from comparisons.
The benchmarks selected by Whisper for calculation of USL are shown in Table 6.14-47 through Table 6.14-51. The similarity of a benchmark to the application is characterized by the correlation coefficient, ck, which is calculated by Whisper. Whisper-calculated correlation coefficients can range between zero (no correlation or anticorrelation) and one (perfect linear correlation). Per [12], at least 15 to 20 very highly correlated systems (ck  0.90) or 25 to 40 moderately correlated systems (ck  0.80) should be included in benchmark evaluations. These thresholds are satisfied for all cases.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                Rev. 16, May 2021 Table 6.14 Whisper-Selected Application Benchmarks for ATR LEU Benchmark          ck        Benchmark          ck        Benchmark              ck ieu-comp-therm-002-003 0.9801 leu-comp-therm-028-010 0.9322 leu-comp-therm-028-017    0.9028 leu-comp-therm-025-004 0.9798 leu-comp-therm-028-013 0.9321 heu-sol-therm-038-008    0.8954 leu-comp-therm-022-005 0.9710 leu-comp-therm-028-019 0.9315 heu-sol-therm-025-003    0.8882 leu-comp-therm-025-003 0.9707 leu-comp-therm-028-011 0.9312 leu-comp-therm-028-002    0.8867 leu-comp-therm-022-004 0.9677 heu-sol-therm-025-015  0.9296 heu-sol-therm-038-007    0.8853 leu-comp-therm-022-006 0.9665 heu-sol-therm-025-018  0.9272 heu-sol-therm-038-010    0.8848 leu-comp-therm-022-007 0.9651 heu-sol-therm-025-014  0.9254 leu-comp-therm-025-001    0.8841 leu-comp-therm-024-002 0.9602 heu-sol-therm-025-009  0.9250 leu-comp-therm-028-007    0.8835 leu-comp-therm-022-003 0.9567 leu-comp-therm-022-002 0.9244 leu-comp-therm-028-001    0.8827 heu-sol-therm-038-002  0.9483 heu-sol-therm-025-011  0.9230 leu-comp-therm-028-008    0.8826 heu-sol-therm-038-022  0.9445 leu-comp-therm-028-020 0.9220 heu-sol-therm-025-001    0.8816 leu-comp-therm-028-015 0.9437 heu-sol-therm-025-017  0.9213 heu-sol-therm-025-002    0.8815 heu-sol-therm-038-006  0.9425 heu-sol-therm-025-013  0.9210 heu-sol-therm-038-003    0.8811 heu-sol-therm-038-023  0.9397 heu-sol-therm-025-010  0.9204 heu-sol-therm-025-004    0.8797 heu-sol-therm-025-006  0.9389 leu-comp-therm-028-014 0.9202 heu-sol-therm-038-028    0.8787 heu-sol-therm-025-008  0.9364 heu-sol-therm-025-012  0.9193 heu-sol-therm-025-005    0.8779 heu-sol-therm-025-007  0.9361 leu-comp-therm-028-018 0.9192 leu-comp-therm-028-003    0.8759 leu-comp-therm-025-002 0.9355 leu-comp-therm-028-012 0.9163 leu-comp-therm-028-009    0.8748 leu-comp-therm-028-016 0.9353 heu-sol-therm-025-016  0.9137 leu-comp-therm-028-005    0.8729 heu-sol-therm-038-021  0.9323 heu-sol-therm-038-009  0.9059 leu-sol-therm-004-001    0.8724 Table 6.14 Whisper-Selected Application Benchmarks for MURR LEU Benchmark          ck        Benchmark          ck        Benchmark              ck leu-comp-therm-025-002 0.9602 leu-comp-therm-028-007 0.9281 heu-sol-therm-038-022    0.8988 leu-comp-therm-025-003 0.9579 leu-comp-therm-028-008 0.9280 leu-comp-therm-028-006    0.8942 ieu-comp-therm-002-003 0.9550 leu-comp-therm-022-004 0.9273 heu-sol-therm-038-006    0.8935 leu-comp-therm-024-002 0.9492 leu-comp-therm-028-001 0.9257 heu-sol-therm-038-023    0.8909 leu-comp-therm-025-004 0.9481 leu-comp-therm-028-003 0.9239 heu-sol-therm-038-021    0.8807 leu-comp-therm-028-013 0.9461 leu-comp-therm-028-009 0.9225 heu-sol-therm-025-018    0.8510 leu-comp-therm-028-011 0.9460 leu-comp-therm-022-001 0.9216 leu-comp-therm-010-009    0.8495 leu-comp-therm-028-010 0.9452 leu-comp-therm-022-005 0.9212 heu-sol-therm-025-015    0.8484 leu-comp-therm-022-002 0.9442 leu-comp-therm-028-020 0.9209 heu-sol-therm-025-006    0.8477 leu-comp-therm-025-001 0.9417 leu-comp-therm-028-018 0.9202 heu-sol-therm-025-008    0.8469 leu-comp-therm-028-015 0.9399 leu-comp-therm-028-005 0.9192 heu-sol-therm-025-007    0.8451 leu-comp-therm-022-003 0.9379 leu-comp-therm-028-004 0.9187 heu-sol-therm-025-017    0.8449 leu-comp-therm-028-014 0.9364 leu-comp-therm-022-006 0.9077 heu-sol-therm-025-014    0.8444 leu-comp-therm-028-012 0.9343 leu-comp-therm-022-007 0.9060 heu-sol-therm-038-009    0.8438 leu-comp-therm-028-016 0.9331 leu-comp-therm-024-001 0.9057 leu-comp-therm-010-010    0.8405 leu-comp-therm-028-002 0.9305 leu-comp-therm-028-017 0.9048 heu-sol-therm-025-016    0.8389 leu-comp-therm-028-019 0.9303 heu-sol-therm-038-002  0.9016 heu-sol-therm-025-013    0.8383 6-293
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                Rev. 16, May 2021 Table 6.14 Whisper-Selected Application Benchmarks for MIT LEU Benchmark          ck        Benchmark          ck        Benchmark            ck leu-comp-therm-024-002 0.9849 heu-sol-therm-025-008  0.9292 leu-comp-therm-028-008  0.9032 leu-comp-therm-022-004 0.9829 heu-sol-therm-025-014  0.9279 heu-sol-therm-038-003    0.9024 leu-comp-therm-025-003 0.9828 heu-sol-therm-025-017  0.9276 heu-sol-therm-038-028    0.9018 leu-comp-therm-022-003 0.9828 leu-comp-therm-028-016 0.9270 leu-comp-therm-028-005  0.8991 leu-comp-therm-025-004 0.9816 heu-sol-therm-025-009  0.9254 leu-comp-therm-010-009  0.8984 leu-comp-therm-022-005 0.9798 heu-sol-therm-025-011  0.9236 leu-comp-therm-028-003  0.8969 leu-comp-therm-022-002 0.9642 heu-sol-therm-025-016  0.9234 leu-comp-therm-007-010  0.8964 ieu-comp-therm-002-003 0.9634 heu-sol-therm-025-010  0.9233 leu-comp-therm-010-010  0.8961 leu-comp-therm-025-002 0.9610 heu-sol-therm-038-009  0.9231 leu-comp-therm-007-007  0.8955 leu-comp-therm-022-006 0.9595 leu-comp-therm-028-019 0.9220 heu-sol-therm-025-003    0.8951 leu-comp-therm-022-007 0.9569 heu-sol-therm-025-012  0.9211 leu-comp-therm-010-011  0.8947 heu-sol-therm-038-002  0.9532 heu-sol-therm-025-013  0.9195 heu-sol-therm-025-002    0.8927 heu-sol-therm-038-022  0.9506 leu-comp-therm-025-001 0.9188 heu-sol-therm-025-001    0.8925 heu-sol-therm-038-006  0.9492 leu-comp-therm-028-014 0.9187 heu-sol-therm-038-027    0.8919 leu-comp-therm-028-010 0.9476 leu-comp-therm-022-001 0.9160 heu-sol-therm-025-005    0.8918 heu-sol-therm-038-023  0.9475 heu-sol-therm-038-008  0.9148 heu-sol-therm-025-004    0.8913 heu-sol-therm-038-021  0.9418 leu-comp-therm-028-012 0.9145 leu-comp-therm-028-009  0.8910 leu-comp-therm-028-015 0.9409 leu-comp-therm-028-001 0.9124 leu-comp-therm-024-001  0.8873 leu-comp-therm-028-013 0.9398 leu-comp-therm-028-002 0.9103 heu-sol-therm-038-005    0.8865 leu-comp-therm-028-011 0.9397 leu-comp-therm-028-007 0.9089 heu-sol-therm-038-018    0.8865 heu-sol-therm-025-018  0.9319 heu-sol-therm-038-007  0.9066 heu-sol-therm-038-025    0.8865 heu-sol-therm-025-006  0.9312 heu-sol-therm-038-010  0.9066 heu-sol-therm-038-026    0.8862 heu-sol-therm-025-007  0.9311 leu-comp-therm-028-020 0.9062 heu-sol-therm-025-015  0.9310 leu-comp-therm-028-018 0.9045 6-294
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                Rev. 16, May 2021 Table 6.14 Whisper-Selected Application Benchmarks for NBSR LEU Benchmark          ck        Benchmark          ck        Benchmark              ck ieu-comp-therm-002-003 0.9754 heu-sol-therm-025-018  0.9211 heu-sol-therm-038-008    0.8871 leu-comp-therm-025-004 0.9751 leu-comp-therm-028-019 0.9207 heu-sol-therm-025-001    0.8846 leu-comp-therm-022-005 0.9662 heu-sol-therm-025-014  0.9206 heu-sol-therm-025-002    0.8844 leu-comp-therm-022-006 0.9660 heu-sol-therm-025-011  0.9199 heu-sol-therm-025-004    0.8824 leu-comp-therm-022-007 0.9647 heu-sol-therm-025-010  0.9179 heu-sol-therm-025-005    0.8795 leu-comp-therm-025-003 0.9602 heu-sol-therm-025-013  0.9159 heu-sol-therm-038-007    0.8784 leu-comp-therm-022-004 0.9599 leu-comp-therm-028-010 0.9156 heu-sol-therm-038-010    0.8765 heu-sol-therm-038-002  0.9444 heu-sol-therm-025-012  0.9151 heu-sol-therm-038-003    0.8754 leu-comp-therm-024-002 0.9440 heu-sol-therm-025-017  0.9151 heu-sol-therm-038-028    0.8707 leu-comp-therm-022-003 0.9423 leu-comp-therm-028-013 0.9149 leu-sol-therm-004-001    0.8655 heu-sol-therm-038-022  0.9400 leu-comp-therm-025-002 0.9147 heu-comp-therm-002-008    0.8624 heu-sol-therm-025-006  0.9398 leu-comp-therm-028-011 0.9140 heu-comp-therm-002-007    0.8615 heu-sol-therm-038-006  0.9382 leu-comp-therm-028-020 0.9115 leu-comp-therm-028-002    0.8608 heu-sol-therm-025-007  0.9361 leu-comp-therm-028-018 0.9081 leu-sol-therm-004-002    0.8595 heu-sol-therm-025-008  0.9356 heu-sol-therm-025-016  0.9069 heu-comp-therm-002-009    0.8592 heu-sol-therm-038-023  0.9347 leu-comp-therm-028-014 0.9022 heu-comp-therm-002-014    0.8591 leu-comp-therm-028-015 0.9336 leu-comp-therm-022-002 0.8991 heu-comp-therm-002-006    0.8588 heu-sol-therm-038-021  0.9277 leu-comp-therm-028-012 0.8983 heu-sol-therm-043-002    0.8582 heu-sol-therm-025-015  0.9252 heu-sol-therm-038-009  0.8966 leu-comp-therm-028-001    0.8571 leu-comp-therm-028-016 0.9249 leu-comp-therm-028-017 0.8917 leu-comp-therm-028-007    0.8568 heu-sol-therm-025-009  0.9243 heu-sol-therm-025-003  0.8888 Table 6.14 Whisper-Selected Application Benchmarks for Air Transport Benchmark          ck        Benchmark          ck        Benchmark              ck heu-sol-therm-009-003  0.9956 heu-sol-therm-019-002  0.9844 heu-sol-therm-019-003    0.9776 heu-sol-therm-001-008  0.9937 heu-sol-therm-009-001  0.9836 heu-comp-therm-002-012    0.9765 heu-sol-therm-001-007  0.9936 heu-sol-therm-001-005  0.9826 heu-comp-therm-002-004    0.9755 heu-sol-therm-043-001  0.9933 heu-comp-therm-002-011 0.9819 heu-comp-therm-002-020    0.9745 heu-sol-therm-001-003  0.9933 heu-comp-therm-002-002 0.9816 heu-sol-therm-025-004    0.9713 heu-sol-therm-001-004  0.9931 heu-sol-therm-050-004  0.9812 heu-sol-therm-025-001    0.9711 heu-sol-therm-001-002  0.9931 heu-sol-therm-050-006  0.9810 heu-sol-therm-025-002    0.9709 heu-sol-therm-001-001  0.9931 heu-sol-therm-050-002  0.9808 heu-comp-therm-002-013    0.9692 heu-sol-therm-050-005  0.9925 heu-sol-therm-050-008  0.9802 heu-comp-therm-002-005    0.9687 heu-sol-therm-009-002  0.9914 heu-sol-therm-050-001  0.9799 heu-sol-therm-038-011    0.9687 heu-sol-therm-010-001  0.9906 heu-comp-therm-002-024 0.9795 heu-sol-therm-038-012    0.9684 heu-sol-therm-050-011  0.9903 heu-comp-therm-002-001 0.9792 heu-sol-therm-038-001    0.9683 heu-comp-therm-002-018 0.9886 heu-sol-therm-050-010  0.9791 heu-comp-therm-002-025    0.9683 heu-comp-therm-002-023 0.9885 heu-comp-therm-002-003 0.9789 heu-sol-therm-038-004    0.9669 heu-sol-therm-019-001  0.9880 heu-sol-therm-025-005  0.9785 heu-sol-therm-038-005    0.9667 heu-comp-therm-002-019 0.9854 heu-sol-therm-011-002  0.9785 heu-sol-therm-001-006  0.9848 heu-sol-therm-011-001  0.9778 6-295
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                    Rev. 16, May 2021 6.14.6.2 Bias Determination The USL is calculated based on the calculational margin, CM, and margin of subcriticality, MOS.
1 The calculational margin and margin of subcriticality are calculated using the following equations:
    -    Bias represents the systematic difference between calculations and benchmark experiments.
    -    Bias uncertainty relates to the uncertainties in the calculations and benchmark experiments.
    -    non-coverage is a non-coverage penalty which may be applied to the calculational margin if all selected benchmarks are too dissimilar to the application.
    -    MOSsoftware is additional margin to address the impact of possible MCNP software errors.
    -    MOSdata is additional margin for uncertainties in nuclear cross-section data.
    -    MOSapplication is additional margin, as set by the analyst using expert judgment, to address other uncertainties or sources of error.
The Bias, Bias uncertainty, non-coverage, and MOSdata are calculated by Whisper based on the selected benchmarks. The MOSsoftware has been set to 0.005 based on the expert opinion of the software developers [11]. The MOSapplication is set to 0.05, which is the administrative margin traditionally used for transportation package criticality analysis.
Whisper results are shown in Table 6.14-52.
Table 6.14 Whisper Results ATR          MURR              MIT          NBSR            Air Parameter LEU          LEU            LEU            LEU        Transport Bias                        0.01096      0.01367          0.01123        0.01036        0.01399 Bias uncertainty            0.00878      0.00788          0.00902        0.00873        0.02042 non-coverage                0.00000      0.00000          0.00000        0.00000        0.00000 MOSsoftware                  0.00500      0.00500          0.00500        0.00500        0.00500 MOSdata                      0.00214      0.00274          0.00187        0.00243        0.00191 MOSapplication              0.05000      0.05000          0.05000        0.05000        0.05000 USL                          0.92312      0.92071          0.92288        0.92348        0.90868 6-296
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                Rev. 16, May 2021 6.14.7 References
: 1. ASTM A269, Standard Specification for Seamless and Welded Austenitic Stainless Steel Tubing for General Service, ASTM International, Revision 2a
: 2. ASTM A554, Standard Specification for Welded Stainless Steel Mechanical Tubing, ASTM International, Revision 3
: 3. LA-UR-17-22234, The Intrinsic Source Constructor Package: Installation and Use, Los Alamos National Laboratory, March 2017
: 4. LA-UR-17-20709, Listing of Available ACE Data Tables, Los Alamos National Laboratory, October 2017
: 5. PNNL-15870, Compendium of Material Composition Data for Radiation Transport Modeling, Pacific Northwest National Laboratory, Revision 1, March 2011
: 6. LA-UR-17-29981, MCNP User's Manual: Code Version 6.2, Los Alamos National Laboratory, October 2017, RSICC Package ID C850MNYCP00
: 7. LA-UR-17-20567, User Manual for Whisper-1.1, Los Alamos National Laboratory, January 2017
: 8. Title 10, "Energy", Code of Federal Regulations, Part 71, Packaging and Transportation of Radioactive Material
: 9. RFP-936, Calculated Critical Radii of Spheres of Plutonium 239 and Uranium 233 with Various Spherical Reflectors, The Dow Chemical Company, June 1967
: 10. LA-UR-17-24260, Release of MCNP6.2 & Whisper-1.1 - Guidance for NCS Users, Los Alamos National Laboratory, May 2017
: 11. B.C. Kiedrowski, F.B. Brown, et al., Whisper: Sensitivity/Uncertainty-Based Computational Methods and Software for Determining Baseline Upper Subcritical Limits, Retrieved from MCNP Reference Collection (Document No. LA-UR-14-26558)
: 12. NEA/NSC/DOC(95)03, International Handbook of Evaluated Criticality Safety Benchmark Experiments, Nuclear Energy Agency
: 13. B. L. Broadhead et al., Sensitivity- and Uncertainty-Based Criticality Safety Validation Techniques, Nuclear Science and Engineering 146, pages 340-366, March 2004 6-297
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                            Rev. 16, May 2021 6.14.8 Sample Input File The case which results in the maximum ksafe value, the worst-case ATR LEU HAC package array, is shown below.
ATR LEU Fuel Plates c
c ===== Configuration Details c Bag Water Density: 1.0 c Inner Tube Water Density: 0.9 c Outer Shell Water Density: 0.0 c Neoprene Location: Inner Tube c Polyethylene Location: Bag c Fuel Plate Additional Spacing: 0.36 c
c *** Cell Cards ***
c ===== Universe 1: Fuel Plates c Fuel Foils 1 1 -16.5      1 -2 12 -13 15 16                                  imp:n=1 u=1 2 1 -16.5      1 -2 22 -23 25 26                                  imp:n=1 u=1 3 1 -16.5      1 -2 32 -33 35 36                                  imp:n=1 u=1 4 1 -16.5      1 -2 42 -43 45 46                                  imp:n=1 u=1 5 1 -16.5      1 -2 52 -53 55 56                                  imp:n=1 u=1 6 1 -16.5      1 -2 62 -63 65 66                                  imp:n=1 u=1 7 1 -16.5      1 -2 72 -73 75 76                                  imp:n=1 u=1 8 1 -16.5      1 -2 82 -83 85 86                                  imp:n=1 u=1 9 1 -16.5      1 -2 92 -93 95 96                                  imp:n=1 u=1 10 1 -16.5    1 -2 102 -103 105 106                              imp:n=1 u=1 11 1 -16.5    1 -2 112 -113 115 116                              imp:n=1 u=1 12 1 -16.5    1 -2 122 -123 125 126                              imp:n=1 u=1 13 1 -16.5    1 -2 132 -133 135 136                              imp:n=1 u=1 14 1 -16.5    1 -2 142 -143 145 146                              imp:n=1 u=1 15 1 -16.5    1 -2 152 -153 155 156                              imp:n=1 u=1 16 1 -16.5    1 -2 162 -163 165 166                              imp:n=1 u=1 17 1 -16.5    1 -2 172 -173 175 176                              imp:n=1 u=1 18 1 -16.5    1 -2 182 -183 185 186                              imp:n=1 u=1 19 1 -16.5    1 -2 192 -193 195 196                              imp:n=1 u=1 c
c Cladding 101 2 -2.7    11 -14 17 18 (-1:2:-12 :13 :-15 :-16)                  imp:n=1 u=1 102 2 -2.7    21 -24 27 28 (-1:2:-22 :23 :-25 :-26)                  imp:n=1 u=1 103 2 -2.7    31 -34 37 38 (-1:2:-32 :33 :-35 :-36)                  imp:n=1 u=1 104 2 -2.7    41 -44 47 48 (-1:2:-42 :43 :-45 :-46)                  imp:n=1 u=1 105 2 -2.7    51 -54 57 58 (-1:2:-52 :53 :-55 :-56)                  imp:n=1 u=1 106 2 -2.7    61 -64 67 68 (-1:2:-62 :63 :-65 :-66)                  imp:n=1 u=1 107 2 -2.7    71 -74 77 78 (-1:2:-72 :73 :-75 :-76)                  imp:n=1 u=1 108 2 -2.7    81 -84 87 88 (-1:2:-82 :83 :-85 :-86)                  imp:n=1 u=1 109 2 -2.7    91 -94 97 98 (-1:2:-92 :93 :-95 :-96)                  imp:n=1 u=1 110 2 -2.7    101 -104 107 108 (-1:2:-102:103:-105:-106)              imp:n=1 u=1 111 2 -2.7    111 -114 117 118 (-1:2:-112:113:-115:-116)              imp:n=1 u=1 112 2 -2.7    121 -124 127 128 (-1:2:-122:123:-125:-126)              imp:n=1 u=1 113 2 -2.7    131 -134 137 138 (-1:2:-132:133:-135:-136)              imp:n=1 u=1 114 2 -2.7    141 -144 147 148 (-1:2:-142:143:-145:-146)              imp:n=1 u=1 115 2 -2.7    151 -154 157 158 (-1:2:-152:153:-155:-156)              imp:n=1 u=1 116 2 -2.7    161 -164 167 168 (-1:2:-162:163:-165:-166)              imp:n=1 u=1 117 2 -2.7    171 -174 177 178 (-1:2:-172:173:-175:-176)              imp:n=1 u=1 118 2 -2.7    181 -184 187 188 (-1:2:-182:183:-185:-186)              imp:n=1 u=1 119 2 -2.7    191 -194 197 198 (-1:2:-192:193:-195:-196)              imp:n=1 u=1 c
c Volume between Fuel Plates (i.e. inside bag) 191 6 -0.998 (-11 :14 :-17 :-18) (-21 :24 :-27 :-28) (-31 :34 :-37 :-38)
(-41 :44 :-47 :-48) (-51 :54 :-57 :-58) (-61 :64 :-67 :-68)
(-71 :74 :-77 :-78) (-81 :84 :-87 :-88) (-91 :94 :-97 :-98) 6-298
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                      Rev. 16, May 2021
(-101:104:-107:-108) (-111:114:-117:-118) (-121:124:-127:-128)
(-131:134:-137:-138) (-141:144:-147:-148) (-151:154:-157:-158)
(-161:164:-167:-168) (-171:174:-177:-178) (-181:184:-187:-188)
(-191:194:-197:-198)
((27 28 14 -21):(37 38 24 -31):(47 48 34 -41):(57 58 44 -51):
(67 68 54 -61):(77 78 64 -71):(87 88 74 -81):(97 98 84 -91):
(107 108 94 -101):(117 118 104 -111):(127 128 114 -121):
(137 138 124 -131):(147 148 134 -141):(157 158 144 -151):
(167 168 154 -161):(177 178 164 -171):(187 188 174 -181):
(197 198 184 -191))                                imp:n=1 u=1 c
c Volume around Fuel Plates 192 7 -0.882  -11:(11 -14 (-17:-18)):(14 -24 (-27:-28)):(24 -34 (-37:-38)):
(34 -44 (-47:-48)):(44 -54 (-57:-58)):(54 -64 (-67:-68)):
(64 -74 (-77:-78)):(74 -84 (-87:-88)):(84 -94 (-97:-98)):
(94 -104 (-107:-108)):(104 -114 (-117:-118)):(114 -124 (-127:-128)):
(124 -134 (-137:-138)):(134 -144 (-147:-148)):(144 -154 (-157:-158)):
(154 -164 (-167:-168)):(164 -174 (-177:-178)):(174 -184 (-187:-188)):
(184 -194 (-197:-198)):194                                  imp:n=1 u=1 c
c ===== Container c == Universe 20/30: Fuel Plates Centered in Container 201 0        -201                              fill=1 (100) imp:n=1 u=20 202 4 -8.0  -202 201                                      imp:n=1 u=20 203 0        -203 202 -204.1 -204.2 -204.3 -204.4          imp:n=1 u=20 204 0            203 -204.1 -204.2 -204.3 -204.4          imp:n=1 u=20 205 4 -8.0            204.1: 204.2: 204.3: 204.4          imp:n=1 u=20 306 4 -8.0  -205                                  fill=20 imp:n=1 u=30 307 3 -1.0  -206 205                                      imp:n=1 u=30 308 0            206                                      imp:n=0 u=30 c
c == Universe 21: Fuel Plates Straight-Shifted in Container 211 like 201 but fill=1 (101) u=21 212 like 202 but              u=21 213 like 203 but              u=21 214 like 204 but              u=21 215 like 205 but              u=21 c
c == Universe 22: Fuel Plates Diagonally-Shifted in Container 221 like 201 but fill=1 (102) u=22 222 like 202 but              u=22 223 like 203 but              u=22 224 like 204 but              u=22 225 like 205 but              u=22 c
c == Universe 31: Fuel Plates Straight-Shifted in Container, 0 deg Rotation 316 like 306 but fill=21      u=31 317 like 307 but              u=31 318 like 308 but              u=31 c
c == Universe 32: Fuel Plates Straight-Shifted in Container, 90 deg Rotation 326 like 306 but fill=21 (201) u=32 327 like 307 but              u=32 328 like 308 but              u=32 c
c == Universe 33: Fuel Plates Straight-Shifted in Container, 180 deg Rotation 336 like 306 but fill=21 (202) u=33 337 like 307 but              u=33 338 like 308 but              u=33 c
c == Universe 34: Fuel Plates Straight-Shifted in Container, 270 deg Rotation 346 like 306 but fill=21 (203) u=34 347 like 307 but              u=34 6-299
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                      Rev. 16, May 2021 348 like 308 but              u=34 c
c == Universe 35: Fuel Plates Diagonally-Shifted in Container, 0 deg Rotation 356 like 306 but fill=22      u=35 357 like 307 but              u=35 358 like 308 but              u=35 c
c == Universe 36: Fuel Plates Diagonally-Shifted in Container, 90 deg Rotation 366 like 306 but fill=22 (201) u=36 367 like 307 but              u=36 368 like 308 but              u=36 c
c == Universe 37: Fuel Plates Diagonally-Shifted in Container, 180 deg Rotation 376 like 306 but fill=22 (202) u=37 377 like 307 but              u=37 378 like 308 but              u=37 c
c == Universe 38: Fuel Plates Diagonally-Shifted in Container, 270 deg Rotation 386 like 306 but fill=22 (203) u=38 387 like 307 but              u=38 388 like 308 but              u=38 c
c ===== Array 1001 0        -301                                        imp:n=1 u=4 lat=1 fill=-1:2 -1:2 0:0 38 38 35 35 38 38 35 35 37 37 36 36 37 37 36 36 1002 0        -302                                fill=4 imp:n=1 1003 3 -1.0    -303 302                                    imp:n=1 1004 0              303                                    imp:n=0 c
c ===== Visualization Universes c All materials defined in unused universes for consistent c VisEd material color map between models 9991 1 1 imp:n=1 u=91 9992 2 1 imp:n=1 u=92 9993 3 1 imp:n=1 u=93 9994 4 1 imp:n=1 u=94 9995 5 1 imp:n=1 u=95 9996 6 1 imp:n=1 u=96 9997 7 1 imp:n=1 u=97 c *** Surface Cards ***
c ===== Fuel Plates 1 pz -61.2775 $ Bottom of fuel meat 2 pz 61.2775 $ Top of fuel meat c
c Fuel Plate Inner Radius 11 1 cz 7.6441 21 2 cz 8.0480 31 3 cz 8.3731 41 4 cz 8.6957 51 5 cz 9.0208 61 6 cz 9.3459 71 7 cz 9.6711 81 8 cz 9.9962 91 9 cz 10.3213 101 10 cz 10.6464 111 11 cz 10.9715 121 12 cz 11.2967 131 13 cz 11.6218 6-300
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report        Rev. 16, May 2021 141 14 cz 11.9444 151 15 cz 12.2695 161 16 cz 12.5946 171 17 cz 12.9197 181 18 cz 13.2448 191 19 cz 13.5674 c
c Fuel Foil Inner Radius 12 1  cz 7.7343 22 2  cz 8.0937 32 3  cz 8.4188 42 4  cz 8.7376 52 5  cz 9.0627 62 6  cz 9.3878 72 7  cz 9.7130 82 8  cz 10.0381 92 9  cz 10.3632 102 10 cz 10.6883 112 11 cz 11.0134 122 12 cz 11.3386 132 13 cz 11.6637 142 14 cz 11.9863 152 15 cz 12.3114 162 16 cz 12.6403 172 17 cz 12.9718 182 18 cz 13.2969 192 19 cz 13.6830 c
c Fuel Foil Outer Radius 13 1  cz 7.7546 23 2  cz 8.1267 33 3  cz 8.4519 43 4  cz 8.7782 53 5  cz 9.1034 63 6  cz 9.4285 73 7  cz 9.7536 83 8  cz 10.0787 93 9  cz 10.4038 103 10 cz 10.7290 113 11 cz 11.0541 123 12 cz 11.3792 133 13 cz 11.7043 143 14 cz 12.0269 153 15 cz 12.3520 163 16 cz 12.6733 173 17 cz 12.9921 183 18 cz 13.3172 193 19 cz 13.7033 c
c Fuel Plate Outer Radius 14 1  cz 7.8448 24 2  cz 8.1724 34 3  cz 8.4976 44 4  cz 8.8201 54 5  cz 9.1453 64 6  cz 9.4704 74 7  cz 9.7955 84 8  cz 10.1206 94 9  cz 10.4458 104 10 cz 10.7709 114 11 cz 11.0960 124 12 cz 11.4211 134 13 cz 11.7462 6-301
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report              Rev. 16, May 2021 144 14 cz 12.0688 154 15 cz 12.3939 164 16 cz 12.7190 174 17 cz 13.0442 184 18 cz 13.3693 194 19 cz 13.8189 c
c Fuel Foil Arcs 15 1  p 0 0 0 0  0 1  0.2983 0.9545 0 16 1  p 0 0 0 0  0 1 -0.2983 0.9545 0 25 2  p 0 0 0 0  0 1  0.3021 0.9533 0 26 2  p 0 0 0 0  0 1 -0.3021 0.9533 0 35 3  p 0 0 0 0  0 1  0.3052 0.9523 0 36 3  p 0 0 0 0  0 1 -0.3052 0.9523 0 45 4  p 0 0 0 0  0 1  0.3081 0.9513 0 46 4  p 0 0 0 0  0 1 -0.3081 0.9513 0 55 5  p 0 0 0 0  0 1  0.3109 0.9504 0 56 5  p 0 0 0 0  0 1 -0.3109 0.9504 0 65 6  p 0 0 0 0  0 1  0.3133 0.9496 0 66 6  p 0 0 0 0  0 1 -0.3133 0.9496 0 75 7  p 0 0 0 0  0 1  0.3157 0.9488 0 76 7  p 0 0 0 0  0 1 -0.3157 0.9488 0 85 8  p 0 0 0 0  0 1  0.3179 0.9481 0 86 8  p 0 0 0 0  0 1 -0.3179 0.9481 0 95 9  p 0 0 0 0  0 1  0.3200 0.9474 0 96 9  p 0 0 0 0  0 1 -0.3200 0.9474 0 105 10 p 0 0 0 0  0 1  0.3218 0.9468 0 106 10 p 0 0 0 0  0 1 -0.3218 0.9468 0 115 11 p 0 0 0 0  0 1  0.3236 0.9462 0 116 11 p 0 0 0 0  0 1 -0.3236 0.9462 0 125 12 p 0 0 0 0  0 1  0.3254 0.9456 0 126 12 p 0 0 0 0  0 1 -0.3254 0.9456 0 135 13 p 0 0 0 0  0 1  0.3270 0.9450 0 136 13 p 0 0 0 0  0 1 -0.3270 0.9450 0 145 14 p 0 0 0 0  0 1  0.3286 0.9445 0 146 14 p 0 0 0 0  0 1 -0.3286 0.9445 0 155 15 p 0 0 0 0  0 1  0.3300 0.9440 0 156 15 p 0 0 0 0  0 1 -0.3300 0.9440 0 165 16 p 0 0 0 0  0 1  0.3314 0.9435 0 166 16 p 0 0 0 0  0 1 -0.3314 0.9435 0 175 17 p 0 0 0 0  0 1  0.3326 0.9431 0 176 17 p 0 0 0 0  0 1 -0.3326 0.9431 0 185 18 p 0 0 0 0  0 1  0.3339 0.9426 0 186 18 p 0 0 0 0  0 1 -0.3339 0.9426 0 195 19 p 0 0 0 0  0 1  0.3352 0.9421 0 196 19 p 0 0 0 0  0 1 -0.3352 0.9421 0 c
c Fuel Plate Arcs 17 1  p 0 0 0 0  0 1  0.3418 0.9398 0 18 1  p 0 0 0 0  0 1 -0.3418 0.9398 0 27 2  p 0 0 0 0  0 1  0.3348 0.9423 0 28 2  p 0 0 0 0  0 1 -0.3348 0.9423 0 37 3  p 0 0 0 0  0 1  0.3366 0.9417 0 38 3  p 0 0 0 0  0 1 -0.3366 0.9417 0 47 4  p 0 0 0 0  0 1  0.3383 0.9410 0 48 4  p 0 0 0 0  0 1 -0.3383 0.9410 0 57 5  p 0 0 0 0  0 1  0.3400 0.9404 0 58 5  p 0 0 0 0  0 1 -0.3400 0.9404 0 67 6  p 0 0 0 0  0 1  0.3414 0.9399 0 68 6  p 0 0 0 0  0 1 -0.3414 0.9399 0 77 7  p 0 0 0 0  0 1  0.3428 0.9394 0 78 7  p 0 0 0 0  0 1 -0.3428 0.9394 0 87 8  p 0 0 0 0  0 1  0.3441 0.9389 0 6-302
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                    Rev. 16, May 2021 88 8 p 0 0 0 0 0 1    -0.3441 0.9389 0 97 9 p 0 0 0 0 0 1    0.3454 0.9385 0 98 9 p 0 0 0 0 0 1    -0.3454 0.9385 0 107 10 p 0 0 0 0 0 1    0.3464 0.9381 0 108 10 p 0 0 0 0 0 1  -0.3464 0.9381 0 117 11 p 0 0 0 0 0 1    0.3475 0.9377 0 118 11 p 0 0 0 0 0 1  -0.3475 0.9377 0 127 12 p 0 0 0 0 0 1    0.3485 0.9373 0 128 12 p 0 0 0 0 0 1  -0.3485 0.9373 0 137 13 p 0 0 0 0 0 1    0.3495 0.9369 0 138 13 p 0 0 0 0 0 1  -0.3495 0.9369 0 147 14 p 0 0 0 0 0 1    0.3504 0.9366 0 148 14 p 0 0 0 0 0 1  -0.3504 0.9366 0 157 15 p 0 0 0 0 0 1    0.3513 0.9363 0 158 15 p 0 0 0 0 0 1  -0.3513 0.9363 0 167 16 p 0 0 0 0 0 1    0.3521 0.9360 0 168 16 p 0 0 0 0 0 1  -0.3521 0.9360 0 177 17 p 0 0 0 0 0 1    0.3528 0.9357 0 178 17 p 0 0 0 0 0 1  -0.3528 0.9357 0 187 18 p 0 0 0 0 0 1    0.3536 0.9354 0 188 18 p 0 0 0 0 0 1  -0.3536 0.9354 0 197 19 p 0 0 0 0 0 1    0.3596 0.9331 0 198 19 p 0 0 0 0 0 1  -0.3596 0.9331 0 c
c ===== Container 201 cz 7.3838          $ IR pipe 202 cz 7.6581          $ OR pipe 203 cz 10.1981        $ 1" insulation 204 rpp -9.6032 9.6032 $ Square Tube Inner Surface
        -9.6032 9.6032
        -999    999 205 rpp -10.033 10.033 $ Square Tube Outer Surface
        -10.033 10.033
        -61.278 61.278 206 rpp -40.513 40.513 $ 12" Water Reflection
        -40.513 40.513
        -91.758 91.758 c
c ===== Array 301 rpp -10.03 10.03  $ Array cell
        -10.03 10.03
        -61.277 61.277 302 rpp -30.089 50.149 $ 4x4 Array boundary
        -30.089 50.149
        -61.276 61.276 303 rpp -60.569 80.629 $ 12" Water Reflection
        -60.569 80.629
        -91.759 91.759 c *** Data Cards ***
c ===== Transformations tr1 0 -6.48 0 $ Fuel Plate  1 tr2 0 -6.12 0 $ Fuel Plate  2 tr3 0 -5.76 0 $ Fuel Plate  3 tr4 0 -5.40 0 $ Fuel Plate  4 tr5 0 -5.04 0 $ Fuel Plate  5 tr6 0 -4.68 0 $ Fuel Plate  6 tr7 0 -4.32 0 $ Fuel Plate  7 tr8 0 -3.96 0 $ Fuel Plate  8 tr9 0 -3.60 0 $ Fuel Plate  9 tr10 0 -3.24 0 $ Fuel Plate 10 tr11 0 -2.88 0 $ Fuel Plate 11 tr12 0 -2.52 0 $ Fuel Plate 12 6-303
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                      Rev. 16, May 2021 tr13 0 -2.16 0 $ Fuel Plate 13 tr14 0 -1.80 0 $ Fuel Plate 14 tr15 0 -1.44 0 $ Fuel Plate 15 tr16 0 -1.08 0 $ Fuel Plate 16 tr17 0 -0.72 0 $ Fuel Plate 17 tr18 0 -0.36 0 $ Fuel Plate 18 tr19 0 0 0 $ Fuel Plate 19 c
tr100 0      -7.4686 0  $ Fuel Plates Centered in Container tr101 0      -7.5    0 $ Fuel Plates Straight-Shifted in Container
*tr102 5.3033 -5.3033 0  $ Fuel Plates Diagonally-Shifted in Container 45 45 90  135 45 90 c
*tr201 0 0 0  90  0 90 180 90 90 $ 90 deg Rotation CCW
*tr202 0 0 0 180 90 90 -90 180 90 $ 180 deg Rotation CCW
*tr203 0 0 0 -90 180 90    0 -90 90 $ 270 deg Rotation CCW c ===== Materials c ===============================================
c U-10Mo Fuel c Density = 16.5 g/cm^3 c ===============================================
m1    42092  -1.273450e-02 42094  -8.131011e-03 42095  -1.415610e-02 42096  -1.500685e-02 42097  -8.690852e-03 42098  -2.221755e-02 42100  -9.063129e-03 92232  -1.820000e-09 92234  -2.366000e-03 92235  -1.815450e-01 92236  -4.186000e-03 92238  -7.219030e-01 c ===============================================
c Aluminum 6061-O c Density = 2.7 g/cm^3 c ===============================================
mt2 al27 m2    12024  -7.794921e-03 12025  -1.028000e-03 12026  -1.176980e-03 13027  -9.719903e-01 14028  -5.511934e-03 14029  -2.900148e-04 14030  -1.979911e-04 22046  -6.969614e-05 22047  -6.421984e-05 22048  -6.498300e-04 22049  -4.868279e-05 22050  -4.756247e-05 24050  -8.138607e-05 24052  -1.632121e-03 24053  -1.886331e-04 24054  -4.784012e-05 25055  -8.799912e-04 26054  -2.309010e-04 26056  -3.758735e-03 26057  -8.835819e-05 26058  -1.196495e-05 29063  -1.883159e-03 29065  -8.668135e-04 30064  -6.887649e-04 30066  -4.116611e-04 6-304
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                    Rev. 16, May 2021 30067  -6.127886e-05 30068  -2.884319e-04 30070  -9.848649e-06 c ===============================================
c Water, Liquid c Density = 1.0 g/cm^3 c ===============================================
mt3 lwtr m3      1001  -1.118683E-01 1002  -2.571290E-05 8016  -8.877466E-01 8017  -3.593967E-04 c ===============================================
c Steel, Stainless 304 c Density = 8.0 g/cm^3 c ===============================================
mt4 fe56 m4      6000  -3.999960e-04 14028  -4.593278e-03 14029  -2.416790e-04 14030  -1.649926e-04 15031  -2.299977e-04 16032  -1.420716e-04 16033  -1.156799e-06 16034  -6.753296e-06 16036  -1.682534e-08 24050  -7.929925e-03 24052  -1.590272e-01 24053  -1.837963e-02 24054  -4.661345e-03 25055  -9.999900e-03 26054  -3.961618e-02 26056  -6.448942e-01 26057  -1.515980e-02 26058  -2.052851e-03 28058  -6.215726e-02 28060  -2.476752e-02 28061  -1.094596e-03 28062  -3.547175e-03 28064  -9.325298e-04 c ===============================================
c Durablanket S c Density = 0.096 g/cm^3 c ===============================================
m5      8016  -5.012956E-01 8017  -2.029452E-04 13027  -2.647353E-01 14028  -2.147528E-01 14029  -1.129939E-02 14030  -7.714015E-03 c ===============================================
c Homogenized Moderator #1 c ===============================================
mt6 lwtr m6      1001  -1.037713e+03 1002  -2.385180e-01 6000  -1.712568e+02 8016  -8.006870e+03 8017  -3.241514e+00 c ===============================================
c Homogenized Moderator #2 c ===============================================
mt7 lwtr 6-305
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report        Rev. 16, May 2021 m7      1001  -9.097014e+02 1002  -2.090946e-01 6000  -6.848088e+02 8016  -6.649161e+03 8017  -2.691857e+00 c
c ===== Source and Run Setup mode n kcode 10000 1.0 50 250 sdef x=d1 y=d2 z=d3 si1 H -27 47 sp1 D 0 1 si2 H -27 47 sp2 D 0 1 si3 H -61.2775 61.2775 sp3 D 0 1 c
prdmp j 50 0 1 50 6-306
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 7.0 PACKAGE OPERATIONS This section provides general instructions for loading and unloading operations of the ATR FFSC. Due to the low specific activity of neutron and gamma emitting radionuclides, dose rates from the contents of the package are minimal. As a result of the low dose rates, there are no special handling requirements for radiation protection.
Package loading and unloading operations shall be performed using detailed written procedures.
The operating procedures developed by the user for the loading and unloading activities shall be performed in accordance with the procedural requirements identified in the following sections.
The closure handle must be rendered inoperable for lifting and tiedown during transport per 10 CFR &sect;71.45. To satisfy this requirement either the closure handle may be removed or the cover installed. If the closure handle cover is utilized it may be stored with the closure assembly in the installed position. When stored with the closure assembly the cover must be removed prior to the package loading and unloading operations and may be reinstalled following installation of the closure. The installation of the closure handle cover is presented in Section 7.1.8, Preparation for Transport.
7.1 Package Loading 7.1.1 Preparation for Loading Prior to loading the ATR FFSC, the packaging is inspected to ensure that it is in unimpaired physical condition. The packaging is inspected for:
Damage to the closure locking mechanism including the spring. Inspect for missing hardware and verify the locking pins freely engage/disengage with the package body mating features.
Damage to the closure lugs and interfacing body lugs. Inspect lugs for damage that precludes free engagement of the closure with the body.
Deformation of the inner shell (payload cavity) that precludes free entry/removal of the payload.
Deformed threads or other damage to the fasteners or body of the loose fuel plate basket.
Damage to the spring plunger, or ball lock pins and end spacers, as applicable, or body of the fuel handling enclosure.
Acceptance criteria and detailed loading procedures derived from this section are specified in user written procedures. These user procedures are specific to the authorized content of the package and inspections ensure the packaging complies with Appendix 1.3.2, Packaging General Arrangement Drawings.
Defects that require repair shall be corrected prior to shipping in accordance with approved procedures consistent with the quality program in effect.
7-1
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 7.1.2 Loading of Contents - ATR HEU Fuel Element Note: This section applies to the ATR HEU fuel element ONLY and the fuel handling enclosure depicted on SAR drawing 60501-30, as summarized in Table 7.1-1. For the ATR LEU fuel element, also known as the LOWE fuel element, see Section 7.1.6, Loading of Contents - ATR LEU Fuel Element.
: 1. Remove the closure by depressing the spring-loaded pins and rotating the closure 45&#xba; to align the closure locking tabs with the mating cut-outs in the body. Remove the closure from the body.
: 2. Remove the fuel handling enclosure if present in the payload cavity.
: 3. Prior to loading, visually inspect the ATR fuel handling enclosure for damage, corrosion, and missing hardware to ensure compliance with Appendix 1.3.2, Packaging General Arrangement Drawings.
: 4. Open the ATR fuel handling enclosure lid and place a fuel element into the holder with the narrow end of the fuel element facing the bottom side of the fuel handling enclosure. As a property protection precaution, the fuel element may optionally be inserted into a polyethylene bag prior to placement in the fuel handling enclosure. Verify the total mass of polyethylene and any tape used to seal the bag is  100 g.
: a. To open the fuel handling enclosure, release the lid by pulling on the spring plunger located at each end and rotate the lid about the hinged side.
: b. To close the fuel handling enclosure, rotate the lid to the closed position, pull the spring plunger located at each end to allow the lid to fully close, align then release the spring plungers with the receiving holes, gently lift the lid to confirm no movement and that the spring plungers are in the locked position.
: 5. Insert the fuel handling enclosure into the package.
: 6. Depress the package closure spring-loaded pins, insert closure onto package body by aligning the closure locking tabs with the mating cut-outs in the body, and rotate the closure to the locked position. Release the spring-loaded pins so that they engage with the mating holes in the package body. Observe the pins to ensure they are in the locked position as illustrated in Figure 7.1-1. The closure is fully locked when both locking pins are compressing the sleeve between the locking pin handle and the closure body.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                Rev. 16, May 2021 Optional TID    Sleeve Loose Location With Closure Rotated 180&deg; Contact Point                                                                    Unlocked When Disengaged                                                                      Position Sleeve Compressed Between Locking Pin Handle                                                                    Locked And Closure Body                                                                    Position When Engaged TID Figure 7.1 Closure Locking Positions 7.1.3 Loading of Contents - Loose ATR Fuel Plates
: 1. Remove the closure by depressing the spring-loaded pins and rotating the closure 45&#xba; to align the closure locking tabs with the mating cut-outs in the body. Remove the closure from the body.
: 2. Remove the fuel plate basket if present in the payload cavity. The fuel plate basket that pertains to this payload is depicted on SAR drawing 60501-20, as summarized in Table 7.1-1.
: 3. Prior to loading, visually inspect the loose fuel plate basket for damage, corrosion, and missing hardware/fastening devices to ensure compliance with Appendix 1.3.2, Packaging General Arrangement Drawings.
: 4. Open the loose fuel plate basket by removing the 8 wing nut fasteners securing each half of the basket.
: 5. Place the fuel plates into one half of the loose fuel plate basket
: a. Ensure the combined weight of the loose fuel plates and optional dunnage is 20 lbs or less. The loose fuel plates may only be ATR fuel plates.
: b. Ensure the combined fissile mass of the loose fuel plates does not exceed 600 g uranium-235.
: c. Flat and curved fuel plates may not be mixed in the same basket.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                Rev. 16, May 2021
: d. As a property protection precaution, the fuel plates may optionally be inserted into polyethylene bag(s) prior to placement in the fuel plate basket. Verify the total mass of polyethylene and any tape used to seal the bag is  100 g.
: e. Dunnage plates may also be included with the loose fuel plates to reduce any gaps with the basket cavity as a property protection precaution. The dunnage plates may be any aluminum alloy and any size deemed appropriate.
: 6. Close the fuel plate basket and verify the basket fasteners are installed and finger tight.
: a. With one half of the basket loaded, carefully place the second half over the fuel plates and match the fastener holes.
: b. Insert the 8 spade head screws through the holes and secure with corresponding wing nut (washer optional).
: c. Tighten the 8 wing nut fasteners finger tight.
: d. Visually check the 4 hex head screws located in the center of the basket to verify that they have not loosened. In the event the screws appear to be loose, tighten the fasteners to drawing requirements.
: 7. Insert the loose fuel plate basket into the package.
: 8. Depress the package closure spring-loaded pins, insert closure onto package body by aligning the closure locking tabs with the mating cut-outs in the body, and rotate the closure to the locked position. Release the spring-loaded pins so that they engage with the mating holes in the package body. Observe the pins to ensure they are in the locked position as illustrated in Figure 7.1-1. The closure is fully locked when both locking pins are compressing the sleeve between the locking pin handle and the closure body.
7.1.4 Loading of Contents - MIT HEU, LEU, & DDE; MURR HEU, LEU,
      & DDE; RINSC, or Cobra Fuel Elements The loading of MIT, MURR, RINSC, and Cobra fuel elements is procedurally identical, except Cobra fuel has one additional step as shown below.
: 1. Remove the closure by depressing the spring-loaded pins and rotating the closure 45&#xba; to align the closure locking tabs with the mating cut-outs in the body. Remove the closure from the body.
: 2. Remove the fuel handling enclosure if present in the payload cavity.
: 3. Prior to loading, visually inspect the fuel handling enclosure for damage, corrosion, and missing hardware to ensure compliance with Appendix 1.3.2, Packaging General Arrangement Drawings.
: 4. Open (disassemble) the fuel handling enclosure and place a fuel element into one enclosure half. Ensure that the MIT, MURR, RINSC, or Cobra fuel element is only used with the corresponding MIT, MURR, RINSC, or Cobra fuel handling enclosure, as summarized in Table 7.1-1. As a property protection precaution, the fuel element may optionally be inserted into a polyethylene bag prior to placement in the fuel handling enclosure. Verify the total 7-4
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                Rev. 16, May 2021 mass of polyethylene and any tape used to seal the bag is limited to the amount shown in the third column of Table 7.1-1.
: a. To open the fuel handling enclosure, remove the two ball lock pins securing each end spacer. Slide each end spacer from the center enclosure halves allowing the enclosure halves to freely come apart.
: b. To close the fuel handling enclosure, with one enclosure half loaded, carefully place the second enclosure half over the fuel element and align the circular ends. Slide one end spacer over the circular end and insert the ball lock pin through the end spacer and enclosure halve alignment holes. Ensure the ball lock pin is in the locked position by observing the pin and locking mechanism protruding from the back side.
Repeat with the second end spacer and ensure it is locked in the same manner.
: 5. When loading Cobra fuel, verify that the alignment post in the Cobra FHE is inserted into one of the nominally 15-mm diameter holes in the end fittings of the Cobra fuel element.
: 6. Insert the fuel handling enclosure into the package.
: 7. Depress the package closure spring-loaded pins, insert closure onto package body by aligning the closure locking tabs with the mating cut-outs in the body, and rotate the closure to the locked position. Release the spring-loaded pins so that they engage with the mating holes in the package body. Observe the pins to ensure they are in the locked position as illustrated in Figure 7.1-1. The closure is fully locked when both locking pins are compressing the sleeve between the locking pin handle and the closure body.
7.1.5 Loading of Contents - Small Quantity Payloads (except RINSC)
The loading of small quantity payloads is procedurally identical.
: 1. Remove the closure by depressing the spring-loaded pins and rotating the closure 45&#xba; to align the closure locking tabs with the mating cut-outs in the body. Remove the closure from the body.
: 2. Remove the fuel handling enclosure if present in the payload cavity.
: 3. Prior to loading, visually inspect the fuel handling enclosure for damage, corrosion, and missing hardware to ensure compliance with Appendix 1.3.2, Packaging General Arrangement Drawings.
: 4. Open (disassemble) the small quantity fuel handling enclosure and place the payload into one enclosure half. The small quantity fuel handling enclosure is depicted on SAR drawing 60501-70, as summarized in Table 7.1-1.
: a. To open the fuel handling enclosure, remove the two ball lock pins securing each end spacer. Slide each end spacer from the center enclosure halves allowing the enclosure halves to freely come apart.
: b. To close the fuel handling enclosure, with one enclosure half loaded, carefully place the second enclosure half over the fuel element, loose fuel plates, or foils and align the circular ends. Slide one end spacer over the circular end and insert the ball lock pin through the end spacer and enclosure halve alignment holes. Ensure the ball lock pin is in the locked position by observing the pin and locking mechanism protruding 7-5
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                Rev. 16, May 2021 from the back side. Repeat with the second end spacer and ensure it is locked in the same manner.
: 5. Dunnage shall be used as necessary to reduce the free space between the small quantity payload face and the SQFHE cavity to a maximum of 1/4 inches or less. The dunnage shall be made from sheets or shapes of aluminum, including steel or aluminum fasteners if required, or may be made from cellulosic material such as cardboard. Neoprene rub strips, nominally 1/8 inch thick, may also be used as a property protection precaution. Neoprene rub strips may be used between the SQFHE and the small quantity payloads and/or between the dunnage and the small quantity payloads. The 1/8 inch neoprene rub strips shall not be stacked in more than two layers between the small quantity payload and any interior face of the SQFHE. Kraft paper and polyethylene sheeting may also be used as property protection.
The sum of the mass of all polyethylene and any other plastic materials such as adhesive tape shall not exceed 100g. The sum of the mass of all cellulosic materials (e.g., paper and cardboard) and neoprene shall not exceed 4 kg.
: 6. Verify that the total weight of the loaded SQFHE is 50 lb or less.
: 7. Insert the fuel handling enclosure into the package.
: 8. Depress the package closure spring-loaded pins, insert closure onto package body by aligning the closure locking tabs with the mating cut-outs in the body, and rotate the closure to the locked position. Release the spring-loaded pins so that they engage with the mating holes in the package body. Observe the pins to ensure they are in the locked position as illustrated in Figure 7.1-1. The closure is fully locked when both locking pins are compressing the sleeve between the locking pin handle and the closure body.
7.1.6 Loading of Contents - ATR LEU Fuel Element Note: This section applies to the ATR LEU fuel element ONLY (also known as the LOWE fuel element) and the fuel handling enclosure depicted on SAR drawing 60501-110, as summarized in Table 7.1-1. For the ATR HEU fuel element, see Section 7.1.2, Loading of Contents - ATR HEU Fuel Element.
: 1. Remove the closure by depressing the spring-loaded pins and rotating the closure 45&#xba; to align the closure locking tabs with the mating cut-outs in the body. Remove the closure from the body.
: 2. Remove the fuel handling enclosure if present in the payload cavity.
: 3. Prior to loading, visually inspect the fuel handling enclosure for damage to ensure compliance with Appendix 1.3.2, Packaging General Arrangement Drawings.
: 4. Open (disassemble) the fuel handling enclosure and place the payload into one enclosure half.
: 5. As a property protection measure, the fuel element may optionally be enclosed in a polyethylene bag prior to placement in the fuel handling enclosure. Verify the total mass of polyethylene, any tape used to seal the bag, and any polymeric material used as strapping to hold the two halves of the FHE together is less than or equal to 200 g.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                Rev. 16, May 2021
: 6. Place the second half over the first, aligning the dowel pins. The two halves shall be retained together at four locations. Strapping material may include plastic ties or straps, metal ties or straps, or duct tape.
: 7. Verify that the total weight of the loaded FHE is 50 lb or less.
: 8. Insert the fuel handling enclosure into the package.
: 9. Depress the package closure spring-loaded pins, insert closure onto package body by aligning the closure locking tabs with the mating cut-outs in the body, and rotate the closure to the locked position. Release the spring-loaded pins so that they engage with the mating holes in the package body. Observe the pins to ensure they are in the locked position as illustrated in Figure 7.1-1. The closure is fully locked when both locking pins are compressing the sleeve between the locking pin handle and the closure body.
7.1.7 Loading of Contents - NBSR DDE
: 1. The NBSR DDE does not have a fuel handling enclosure. Prior to loading, protect the NBSR DDE using disposable packing such as cardboard blocking.
: 2. As a property protection measure, the NBSR DDE may optionally be enclosed in a polyethylene bag prior to the addition of packing materials. Verify the total mass of polyethylene, any tape used to seal the bag, and any polymeric material used to hold the packing material together is less than or equal to 200 g, as shown in Table 7.1-1.
: 3. Prepare end blocking so that the NBSR DDE is located approximately in the package cavity center. The maximum mass of all packing/blocking material is 4 kg (8.8 lb).
: 4. Verify that the total weight of the NBSR DDE and packing/blocking is 50 lb or less.
: 5. Remove the package closure by depressing the spring-loaded pins and rotating the closure 45&#xba; to align the closure locking tabs with the mating cut-outs in the body. Remove the closure from the body.
: 6. Insert the payload into the package.
: 7. Depress the package closure spring-loaded pins, insert closure onto package body by aligning the closure locking tabs with the mating cut-outs in the body, and rotate the closure to the locked position. Release the spring-loaded pins so that they engage with the mating holes in the package body. Observe the pins to ensure they are in the locked position as illustrated in Figure 7.1-1. The closure is fully locked when both locking pins are compressing the sleeve between the locking pin handle and the closure body.
7.1.8 Preparation for Transport
: 1. Install the closure handle cover by aligning the cover against the handle and insert the fastener through the holes in the cover and behind the handle as illustrated in Figure 7.1-2.
Once installed, the cover renders the handle inoperable for lifting or tiedown during transport. Option: In lieu of installing the cover, the closure handle may be removed as a method of rendering the handle inoperable for lifting or tiedown during transport.
: 2. Install the tamper indicating device between the posts on the package closure and body.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                              Rev. 16, May 2021
: 3. Perform a survey of the dose rates and levels of non-fixed (removable) radioactive contamination per 49CFR &sect;173.441 and 49CFR &sect;173.443, respectively. The contamination measurements shall be taken in the most appropriate locations to yield a representative assessment of the non-fixed contamination levels.
: 4. Complete the necessary shipping papers in accordance with Subpart C of 49 CFR &sect;172.
: 5. Ensure that the package markings are in accordance with 10 CFR &sect;71.85(c) and Subpart D of 49 CFR &sect;172. Package labeling shall be in accordance with Subpart E of 49CFR &sect;172.
Package placarding, for either single package transport or the racked configuration, shall be in accordance with Subpart F of 49 CFR &sect;172.
: 6. Transfer the package to the conveyance and secure the package(s).
Figure 7.1 Closure Handle Cover Installation 7-8
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                Rev. 16, May 2021 Table 7.1 Fuel Handling Enclosure Usage Summary FHE SAR        Maximum mass of Fuel Element          Section No.        Drawing No.        poly and tape, g ATR HEU                        7.1.2            60501-30                100 ATR Loose Plates                7.1.3            60501-20                100 MIT HEU                        7.1.4            60501-40                100 MIT LEU & DDE                  7.1.4            60501-40                200 MURR HEU                        7.1.4            60501-50                100 MURR LEU & DDE                  7.1.4          60501-111                200 RINSC                          7.1.4            60501-60                100 Cobra HEU & LEU                7.1.4            60501-90                100 Small Quantity                  7.1.5            60501-70                100 ATR LEU                        7.1.6          60501-110                200 NBSR DDE                        7.1.7              NA                    200 7.2 Package Unloading 7.2.1 Receipt of Package from Conveyance Radiation and contamination surveys shall be performed upon receipt of the package and the package shall be inspected for damage as required by and in accordance with the users personnel protection or ALARA program. In addition, the tamper indicating device (TID) shall be inspected. A missing TID or indication of damage to a TID is a Safeguards and Security concern. Disposition of such an incident is beyond the scope of this SAR.
7.2.2 Removal of Contents
: 1. Remove tamper indicating device.
: 2. Remove the package closure by depressing the spring-loaded pins and rotating the closure 45&#xba; to align the closure locking tabs with the mating cut-outs in the body. Remove the closure from the body.
: 3. Remove the payload container.
: 4. Open the payload container (fuel handling enclosure or loose fuel plate basket) and remove the contents.
: a. Open the ATR HEU fuel handling enclosure by releasing the spring plunger located at each end and rotate the lid about the hinged side.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021
: b. Open the loose fuel plate basket by removing the 8 wing nut fasteners securing each half of the basket.
: c. Open the MIT, MURR, RINSC, Cobra, or small quantity payload fuel handling enclosure by removing the two ball lock pins and sliding the end spacers from each end of the enclosure halves.
: d. Open the ATR LEU fuel handling enclosure by removing the straps holding the two halves together.
: 5. Close the fuel handling enclosure lid or loose fuel plate basket as appropriate. If required, return the empty payload container to the package.
: a. To close the ATR HEU fuel handling enclosure, rotate the lid to the closed position, pull the spring plunger located at each end to allow the lid to fully close, align then release the spring plungers with the receiving holes, gently lift the lid to confirm no movement and that the spring plungers are in the locked position.
: b. To close the loose fuel plate basket, place each half of the basket together and align the fastener holes. Insert the 8 spade head screws through the holes and secure with corresponding wing nut (washer optional). Tighten each wing nut finger tight.
: c. To close the MIT, MURR, RINSC, Cobra, or small quantity payload fuel handling enclosure, place each enclosure half together and align the circular ends. Slide one end spacer over the circular end and insert the ball lock pin through the end spacer and enclosure halve alignment holes. Ensure the ball lock pin is in the locked position by observing the pin and locking mechanism protruding from the back side.
Repeat with the second end spacer and ensure it is locked in the same manner.
: d. To close the ATR LEU fuel handling enclosure, align the dowel pins. Apply at least two straps or lengths of duct tape to retain the empty halves together.
: 6. Depress the package closure spring-loaded pins, insert closure onto package body by aligning the closure locking tabs with the mating cut-outs in the body, and rotate the closure to the locked position. Release the spring-loaded pins so that they engage with the mating holes in the package body. Observe the pins to ensure they are in the locked position as illustrated in Figure 7.1-1. The closure is fully locked when both locking pins are compressing the sleeve between the locking pin handle and the closure body.
7.3 Preparation of Empty Package for Transport Empty packages are prepared and transported per the guidelines of 49 CFR &sect;173.428. The packaging is inspected to ensure that it is in an unimpaired condition and is securely closed.
Any labels previously applied in conformance with subpart E of 49CFR &sect;172 are removed, obliterated, or covered and the Empty label prescribed in 49 CFR &sect;172.450 is affixed to the packaging.
7.4 Other Operations This section does not apply.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 8.0 ACCEPTANCE TESTS AND MAINTENANCE PROGRAM 8.1 Acceptance Tests Per the requirements of 10 CFR &sect;71.85, the inspections and tests to be performed prior to first use of the package are described in this section.
8.1.1 Visual Inspections and Measurements All packaging dimensions, tolerances, general notes, materials of construction, and assembly shall be examined in accordance with the requirements delineated on the drawings in Appendix 1.3.2, Packaging General Arrangement Drawings. Source inspections and final release of the packaging will be performed, verifying the quality characteristics were inspected and that the packaging is acceptable. Any characteristic that is out of specification shall be reported and dispositioned in accordance with the quality assurance program in effect.
8.1.1.1 Compression Spring The compression spring is a component of the closure locking system that maintains the locking pin in the closed position. The compression spring shall be procured to Stock Precision Engineered Components (SPEC) catalog number C0360-035-1120 specification, or equivalent, which includes the following:
Material shall be approximately 0.035 inch diameter stainless steel wire.
The nominal outside diameter of the spring shall be approximately 0.36 inches.
The free length of the spring shall be approximately 1.12 inches.
The solid height of the spring shall be approximately 0.33 inches.
The spring shall have a 4.77 (-.1, +.5) lb load at a load length of approximately 0.55 inches.
The spring rate shall be 8.33 (-.1, +.5) lbs/in.
8.1.1.2 Roll Pin The roll pin is a component of the closure locking system that maintains the locking pin in the closed position. The roll pin shall be procured to Stock Drive Products/Sterling Instrument (SDP/SI) catalog number A9Y35-0324 specification, or equivalent, which includes the following:
Material shall be stainless steel.
The free diameter of the roll pin shall be between 0.099 to 0.103 inches.
The length of the roll pin shall be approximately 0.75 inches 8-1
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                      Rev. 16, May 2021 8.1.1.3 Insulating Blanket The ceramic fiber insulating blanket is a component of the body and closure assemblies used to reduce heat transfer during thermal events. The insulating blanket shall be procured to Unifrax Durablanket S 6 lb/ft3 specification, or equivalent, which includes the following:
The material shall be comprised of inorganic ceramic fibers.
The nominal thickness shall be 0.5 (-0, +.2) inches.
The nominal density shall be 6 (-15%, +30%) lb/ft3.
The specific heat shall be 0.25 Btu/lbm-&deg;F minimum.
The thermal conductivity shall be 0.145 Btu/hr-ft-&deg;F or less at 1200&deg;F.
8.1.2 Weld Examinations All welds shall be examined in accordance with the requirements delineated on the drawings in Appendix 1.3.2, Packaging General Arrangement Drawings. Visual examinations are performed in accordance with AWS D1.61, Section 6 for stainless steel, AWS D1.22, for aluminum, and penetrant examinations are performed under procedures written to ASTM E165-02, Standard Test Method for Liquid Penetrant Examination.
8.1.3 Structural and Pressure Tests The packaging does not retain pressure and no pressure testing is required prior to use.
8.1.4 Leakage Tests The packaging contains no seals or containment boundaries that require leakage rate testing.
8.1.5 Component and Material Tests No component or material tests are required for this packaging.
8.1.6 Shielding Tests The packaging does not contain any biological shielding. Shielding tests are not required.
8.1.7 Thermal Tests The material thermal properties utilized in Chapter 3.0, Thermal are nominal. However, the thermal analyses in which these values are used are consistently conservative for the Normal Conditions of Transport (NCT) and Hypothetical Accident Condition (HAC). Therefore, specific acceptance tests for material thermal properties are not required or performed.
1 ANSI/AWS D1.6:1999, Structural Welding Code - Stainless Steel, American Welding Society (AWS).
2 ANSI/AWS D1.2:2003, Structural Welding Code - Aluminum, American Welding Society.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 8.1.8 Miscellaneous Tests No other acceptance tests are necessary for the packaging.
8.2 Maintenance Program This section describes the maintenance program used to ensure continued performance of the packaging. The packaging is maintained consistent with a 10 CFR 71 subpart H QA program.
Packagings that do not conform to the license drawings are removed from service until they are brought back into compliance. Repairs are performed in accordance with approved procedures and consistent with the quality assurance program in effect.
8.2.1 Structural and Pressure Tests There are no structural or pressure tests that are necessary to ensure continued performance of the packaging.
8.2.2 Leakage Rate Tests No leakage rate tests are necessary to ensure continued performance of the packaging.
8.2.3 Component and Material Tests There is no predetermined replacement schedule for any packaging components and there are no items that would be expected to wear or become damaged during normal usage. The items identified in this section are routinely used during operations and shall be visually inspected prior to each use. Damaged components shall be repaired or replaced prior to further use.
8.2.3.1 Packaging Body and Closure The closure assembly locking pin spring shall be visually inspected and replaced if it becomes damaged or otherwise fails to function properly (Drawing 60501-10, Item 20, of Appendix 1.3.2, Packaging General Arrangement Drawings).
The index lug screws and corresponding tap, or optional wire insert, shall be visually inspected for deformed or stripped threads prior to installation of the screws (Drawing 60501-10, Items 3 and 16).
8.2.3.2 ATR Fuel Handling Enclosure The spring plunger shall be visually inspected and replaced if it becomes damaged or otherwise fails to function properly (Drawing 60501-30, Item 6, of Appendix 1.3.2, Packaging General Arrangement Drawings).
8.2.3.3 Loose Fuel Plate Basket All threaded components shall be visually inspected as they are installed for deformed or stripped threads (Drawing 60501-20, Items 2, 3, 4, and 5 of Appendix 1.3.2, Packaging General Arrangement Drawings).
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                              Rev. 16, May 2021 8.2.3.4 Fuel Handling Enclosure The ball lock pin used with the MIT, MURR, RINSC, Small Quantity, and Cobra FHE shall be visually inspected and replaced if it becomes damaged or otherwise fails to function properly, according to the drawings of Appendix 1.3.2, Packaging General Arrangement Drawings:
MIT HEU & LEU FHE, Drawing 60501-40, Item 4 MURR HEU FHE, Drawing 60501-50, Item 4 MURR LEU FHE, Drawing 60501-111, Item 5 RINSC FHE, Drawing 60501-60, Item 5 Small Quantity FHE, Drawing 60501-70, Item 4 Cobra FHE, Drawing 60501-90, Item 4 The ATR LEU FHE (Drawing 60501-110) requires no maintenance.
8.2.4 Thermal Tests No thermal tests are necessary to ensure continued performance of the packaging.
8.2.5 Miscellaneous Tests No miscellaneous tests are required to ensure continued performance of the packaging.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                          Rev. 16, May 2021 9.0 QUALITY ASSURANCE The Advanced Test Reactor Fresh Fuel Shipping Container (ATR FFSC) is anticipated to be used by both U.S. Department of Energy (DOE) and U.S. Nuclear Regulatory Commission (NRC) licensed users. 10 CFR &sect;71.101, Quality assurance requirements, requires each licensees quality assurance program to be approved by the Commission before any use of the package for shipments.
NRC licensed users shall follow their NRC approved quality assurance program and be identified by the Commission as an authorized user. For DOE and its subcontractors, this chapter defines the approved Quality Assurance (QA) requirements and methods of compliance applicable to the ATR FFSC package.
The ATR FFSC package described in this SAR is used to transport unirradiated single fuel elements. The QA requirements for packagings are described in Subpart H of 10 CFR Part 71 (10 CFR 71). Subpart H is an 18-criteria QA program based on ANSI/ASME NQA-1. Guidance for QA programs for packaging is provided by NRC Regulatory Guide 7.101. The DOE QA requirements for the use of 10CFR71 certified packagings are described in DOE Order 460.1B2.
The ATR FFSC packaging is designed and built for Idaho National Laboratory (INL).
Procurement, design, fabrication, assembly, testing, maintenance, repair, modification, and use of the ATR FFSC package are all done under QA programs that meet all applicable NRC and DOE QA requirements.
The DOE Idaho Operations Office approved QA program is implemented for all Nuclear Safety activities. Compliance with NRC and DOT packaging and transportation requirements is mandated by DOE Order 460.1B.
This document establishes the programmatic requirements for site-wide implementation and serves as the basis for INL quality assurance program acceptability. It is designed such that implementation of the full scope of requirements as stated in DOE Orders 414.1C, Quality Assurance and 460.1B Packaging and Transport Safety, constitutes compliance to nuclear safety quality assurance criteria required by 10 CFR 830, Subpart A, Nuclear Safety Management Quality Assurance Requirements.
A detailed discussion of the QA program which governs ATR FFSC packaging operations is presented on the following pages to demonstrate compliance with 10 CFR 71, Subpart H.
9.1 Organization 9.1.1 ATR FFSC Project Organization This section identifies the organizations involved and describes the responsibilities of and interactions between these organizations.
1 U.S. Nuclear Regulatory Commission, Regulatory Guide 7.10, Establishing Quality Assurance Programs for Packaging Used in transport of Radioactive Material, Revision 2, March 2005.
2 U.S. Department of Energy Order 460.1B, Packaging and Transportation Safety, 4-4-03.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                Rev. 16, May 2021 9.1.1.1 Idaho National Laboratory (INL)
INL Contractor Management has overall responsibility for successfully accomplishing activities.
Management provides the necessary planning, organization, direction, control, resources, and support to achieve their defined objectives. Management is responsible for planning, performing, assessing, and improving the work.
INL Contractor Management is responsible for establishing and implementing policies, plans, and procedures that control the quality of work, consistent with requirements.
INL Contractor Management responsibilities include:
Ensuring adequate technical and QA training is provided for personnel performing activities.
Ensuring compliance with all applicable regulations, DOE orders and requirements, and applicable federal, state, and local laws.
Ensuring personnel adhere to procedures for the generation, identification, control, and protection of QA records.
Exercising authority and responsibility to STOP unsatisfactory work such that cost and schedule do not override environmental, safety, or health considerations.
Developing, implementing, and maintaining plans, policies, and procedures that implement the Quality Assurance Program Description (QAPD).
Identifying, investigating, reporting, and correcting quality problems.
Achieving and maintaining quality in their respective areas. (Quality achievement is the responsibility of those performing the work. Quality achievement is verified by persons or organizations not directly responsible for performing the work.)
Empowering employees by delegating authority and decision making to the lowest appropriate level in the organization.
9.1.1.2    Members of the INL Contractor Workforce (at all levels)
Implement the organizations procedures to meet QA requirements.
Comply with administrative and technical work control requirements.
Identify and report issues to the responsible manager for resolution and continuous improvement for the work being performed.
Seek, identify, and recommend work methods or procedural changes that would improve quality and efficiency.
9.1.1.3    INL Contractor Quality Assurance Management The INL Contractor QA Management provides independent oversight of all quality related activities.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 9.2 Quality Assurance Program 9.2.1 General The INL Contractors QA Program defines and establishes requirements for programs, projects, and activities.
The INL Contractor QA program is developed and maintained through an ongoing process that selectively applies QA criteria as appropriate to the function or work activity being performed.
Applicable QA criteria consist of the following:
Title 10 CFR Subpart 71, Packaging and Transportation of Radioactive Material Title 10 CFR 830.120, Quality Assurance Requirements ASME NQA-1-2000, Quality Assurance Requirements for Nuclear Facility Application DOE O 414.1C, Quality Assurance DOE O 461.1B, Packaging and Transport Safety DOE G 414.1-1A, Management Assessment and Independent Assessment The INL Contractor QA Program is inclusive of applicable requirements from criteria noted above and addresses the following for this SAR:
Organization                                      Records Quality Assurance Program                        Work Process Implementation of the QA Program                  Procurement Personnel Qualification and Training              Inspection and Testing Quality Improvement                              Management Assessments Documents                                        Independent Assessment The INL Contractor QA Director is responsible for ensuring implementation of requirements as defined within the QA program and requirements of this SAR, including design, procurement, fabrication, inspection, testing, maintenance, and modifications. Procurement documents are to reflect applicable requirements from 10 CFR 71, Subpart H, ASME NQA-1 and the QA program.
INL Contractor Quality Management assesses the adequacy and effectiveness of the QA program to ensure effective implementation inclusive of objective evidence and independent verification, where appropriate, to demonstrate that specific project and regulatory objectives are achieved.
All INL Contractor personnel and contractors are responsible for effective implementation of the QA program within the scope of their responsibilities. INL Contract packaging and quality engineers are responsible for inspection and testing and are to be qualified, as appropriate, through minimum education and/or experience, formal training, written examination and/or other demonstration of skill and proficiency. Objective evidence of qualifications and capabilities are to be maintained as required. As appropriate, the initial employee training should consist of the following:
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 General employee indoctrination Program indoctrination QA program training Applicable NRC and DOT requirements.
Note: Only packaging engineers and Quality Engineers with training and/or experience in applicable NRC and DOT requirements and Safety Analysis Reports (SARs) can plan or determine the application of internal INL processes to ensure compliance with Chapter 9 and this SAR.
9.2.2 ATR FFSC-Specific Program The ATR FFSC was designed and tested as described in Chapter 2, Structural Evaluation, of this SAR. QA requirements are invoked in the design, procurement, fabrication, assembly, testing, maintenance, and use of the packaging to ensure established standards are maintained. Items and activities to be controlled and documented are described in this chapter.
9.2.3 QA Levels Materials and components of the ATR FFSC are designed, procured, fabricated, assembled, and tested using a graded approach under a 10 CFR 71, Subpart H equivalent QA Program and Regulatory Guide (RG) 7.10. Under that program, the categories critical to safety are established for all ATR FFSC packaging components. These defined quality categories consider the impact to safety if the component were to fail or perform outside design parameters.
9.2.3.1 Graded Quality Category A Items:
These items and services are critical to safe operation and include structures, components, and systems whose failure could directly result in a condition adversely affecting public health and safety. The failure of a single item could cause loss of primary containment leading to a release of radioactive material beyond regulatory requirements, loss of shielding beyond regulatory requirements, or unsafe geometry compromising criticality control.
9.2.3.2 Graded Quality Category B Items:
These items and services have a major impact on safety and include structures, components, and systems whose failure or malfunction could indirectly result in a condition adversely affecting public health and safety. The failure of a Category B item, in conjunction with the failure of an additional item, could result in an unsafe condition.
9.2.3.3 Graded Quality Category C Items:
These items and services have a minor impact on safety and include structures, components, and systems whose failure or malfunction would not significantly reduce the packaging effectiveness and would not be likely to create a situation adversely affecting public health and safety.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 9.2.3.4 Application of Quality Categories The design effort and requirements for a QA program are interrelated and are developed simultaneously. To ensure the development of a QA program in which the application of QA requirements is commensurate with their safety significance, engineering personnel perform a systematic analysis of each component, structure, and system to assess the consequences to the health and safety of the public and the environment that would result from malfunction or failure of such items. This engineering assessment is initiated during the design process and performed in accordance with approved procedures. Establishment of the engineering basis during the design process enables a uniform, consistent application of QA requirements during fabrication, use, and maintenance of packaging.
A logical sequence is established to identifying realistic QA requirements would involve (1) classifying each structure, system, and component (2) grouping items classified as important to safety into quality categories; and (3) specifying the applicable level of QA effort for each category.
The Design Authority (DA) identifies the critical characteristics when they identify design attributes necessary to preserve the safety support function. As necessary, the DA also ensures critical characteristics are included in this SAR by the identification of SSCs and their QA Category designations. Additionally, this SAR includes the safety function, design, and operational attributes necessary for reliable performance. The DA applies design criteria to the design, operation, and maintenance of each critical SSC including recommended codes and standards, as required by RG 7.10. QA requirements shall be applied as necessary to assure the SSCs can perform their function.
The package-specific safety documents identify systems, structures, and components (SSCs) that are important to the safety functions for transportation. As appropriate, the hazard analysis and accident scenarios in the safety basis documents help identify SSCs that must function in order to prevent or mitigate these events. These SSCs are then identified using the classification system found in the NRC QA Category system provided in NRC Regulatory Guide (RG) 7.10. The categories as defined in RG 7.10, and listed below, are analogous to Safety Class, Safety Significant, and General Service that are identified for facility SSCs.
Upon custodianship of the ATR FFSC packages by INL, functional classifications will be used for site operations and activities related to the ATR FFSC. The method of classification is documented as follows.
Quality Category A:
Critical impact on safety and associated functional requirements - items or components whose single failure or malfunction could directly result in an unacceptable condition of containment, shielding, or nuclear criticality control. This is functionally equivalent to safety class designation used for nuclear facility safety.
Quality Category B:
Impact on safety and associated functional requirement - components whose failure or malfunction in conjunction with one other independent failure or malfunction could result in an unacceptable condition of containment, shielding, or nuclear criticality control. This is functionally equivalent to safety significant designation used for nuclear facility safety.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 Quality Category C:
Minor impact on safety and associated functional requirements - components whose failure or malfunction would not result in an unacceptable condition of containment, shielding, or nuclear criticality control regardless of other single failures. This is functionally equivalent to designations given to components that do not meet safety class or safety significant criteria used for nuclear facility safety.
The tabulation of this classification process is provided in Tables 9.2-1 and 9.2-2.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                              Rev. 16, May 2021 Table 9.2 QA Categories for Design and Procurement of ATR FFSC Subcomponents Component                        Subcomponent                      Category Outer Square Tube                        A Inner Round Tube                        A Bottom End Plate                        A Closure End Plate                        A Stiffening Ribs                        A Body Assembly                  Thermal Shield Sheet                        B Insulation                          B Tamper Indicating Device Dowel C
Pin Index Lug Screw                        B Weld Wire                          A Outer Plate, Closure                      A Inner Plate (Insulation Pocket)                B Closure Locking Hardware B
(Pin, Handle, Spring, etc.)
Closure Assembly Insulation                          B Tamper Indicating Device Dowel C
Pin Weld Wire                          A Loose Fuel Plate Basket Machined Aluminum Body                        A (Drawing No. 60501-20) and Fuel Handling Enclosures Drawing Nos. 60501-40,              All Other Components*                      C 60501-50, and 60501-111 Fuel Handling Enclosures Drawing Nos. 60501-30,                  All Components*                        C 60501-60, 60501-70, 60501-90, and 60501-110
*Note: neoprene in all fuel handling enclosures is categorized as not important to safety (NITS).
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                    Rev. 16, May 2021 Table 9.2 Level of Quality Assurance Effort per QA Element 10CFR71                                                                      QA Subpart H                                                                Category QA Level of QA Effort Element                                                                  A    B  C QA Organization (&#xb6;9.1)
Organizational structure and authorities defined        X    X 1
Responsibilities defined                                X    X (71.103)
Reporting levels established                            X    X Management endorsement                                  X    X QA Program (&#xb6;9.2) 2          Implementing procedures in place                        X    X (71.105)      Trained personnel                                        X    X Activities controlled                                    X    X Design (&#xb6;9.3)
Control of design process and inputs                    X    X  X 3          Control of design input                                  X    X  X (71.107)      Software validated and verified                          X    X Design verification controlled                          X    X Quality category assessment performed                    X    X Procurement Document Control (&#xb6;9.4)
Complete traceability                                    X    X 4
Qualified suppliers list                                X    X (71.109)
Commercial grade dedicated items acceptable              X    X Off-the-shelf item                                                X Instructions, Procedures, and Drawings (&#xb6;9.5) 5 Must be written and controlled                          X    X (71.111)
Qualitative or quantitative acceptance criteria          X    X Document Control (&#xb6;9.6) 6          Controlled issuance                                      X    X (71.113)      Controlled changes                                      X    X Procurement documents                                    X    X  X 9-8
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                    Rev. 16, May 2021 10CFR71                                                                                    QA Subpart H                                                                                Category QA Level of QA Effort Element                                                                                A    B  C Control of Purchased Material, Equipment, and Services (&#xb6;9.7)
Source evaluation and selection plans                                X    X Evidence of QA at supplier                                            X    X 7            Inspections at supplier, as applicable                                X    X (71.115)        Receiving inspection                                                  X    X Objective proof that all specifications are met                      X    X Audits/surveillances at supplier facility, as applicable              X    X Incoming inspection for damage only                                            X Identification and Control of Material, Parts, and Components (&#xb6;9.8) 8            Positive identification and traceability of each item                X    X (71.117)        Identification and traceable to heats, lots, or other groupings      X    X Identification to end use drawings, etc.                                      X Control of Special Processes (&#xb6;9.9)
All welding, heat treating, and nondestructive testing done by        X    X 9
qualified personnel (71.119)
Qualification records and training of personnel                      X    X No special processes                                                          X Inspection (&#xb6;9.10)
Documented inspection to all specifications required                  X    X Examination, measurement, or test of material or processed            X    X 10            product to assure quality (71.131)        Process monitoring if quality requires it                            X    X Inspectors must be independent of those performing operations        X    X  X Qualified inspectors only                                            X    X  X Receiving inspection                                                  X    X  X Test Control (&#xb6;9.11)
Written test program                                                  X    X Written test procedures for requirements in the package              X    X 11            approval (71.123)        Documentation of all testing and evaluation                          X    X Representative of buyer observes all supplier acceptance tests if    X specified in procurement documents No physical tests required                                                    X 9-9
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 10CFR71                                                                                  QA Subpart H                                                                              Category QA Level of QA Effort Element                                                                              A    B  C Control of Measuring and Test Equipment (&#xb6;9.12)
Tools, gauges, and instruments to be in a formal calibration          X    X 12 program (71.125)
Only qualified inspectors                                            X    X No test required                                                              X Handling, Storage, and Shipping (&#xb6;9.13) 13 Written plans and procedures required                                X    X (71.127)
Routine handling                                                              X Inspection, Test, and Operating Status (&#xb6;9.14) 14        Individual items identified as to status or condition                X    X (71.129)      Stamps, tags, labels, etc., must clearly show status                  X    X  X Visual examination only                                                        X Nonconforming Materials, Parts, or Components (&#xb6;9.15) 15        Written program to prevent inadvertent use                            X    X  X (71.131)      Nonconformance to be documented and closed                            X    X  X Disposal without records                                                      X 16                          Corrective Action (&#xb6;9.16)
(71.133)      Objective evidence of closure for conditions adverse to quality      X    X  X QA Records (&#xb6;9.17)
Design and use records                                                X    X Results of reviews, inspections, test, audits, surveillance, and      X    X materials analysis 17        Personnel qualifications                                              X    X (71.135)      Records of fabrication, acceptance, and maintenance retained          X    X  X throughout the life of package Record of package use kept for three years after shipment            X    X All records managed by written plans for retention and disposal      X    X Procurement records                                                  X    X  X Audits (&#xb6;9.18) 18 Written plan of periodic audits                                      X    X (71.137)
Lead auditor certified                                                X    X 9-10
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                          Rev. 16, May 2021 9.3 Package Design Control As required by the INL Contractors Quality Program, design processes shall be established and implemented to satisfy the requirements of the QAPD. These requirements are to be in accordance with:
10 CFR 830.122(f), Criterion 6 - Performance/Design3 DOE Order 414C, CRD, Attachment 1, 2.b.(2), Criterion 6 - Design.
Requirements are implemented to ensure processes and procedures are in place to ensure design features of packaging systems are appropriately translated into specifications, drawings, procedures, and instructions. Design control measures are established for criticality, shielding, thermal, and structural analyses under both normal and accident condition analyses as defined in NRC regulations.
The INL Contractor is responsible for maintaining the package and this SAR. The design documents (e.g., drawings and specifications) are controlled by incorporation into this SAR, which will be reviewed and approved by the NRC.
The design of the ATR FFSC was performed under an NRC-approved QA Program as required by INL. Design inputs consist of an INL statement of work, applicable DOE orders, national standards, specifications, and drawings.
Procedures control design activities to ensure the following occur:
Design activities are planned, controlled, and documented.
Regulatory requirements, design requirements, and appropriate quality standards are correctly translated into specifications, drawings, and procedures.
Competent engineering personnel, independent of design activities, perform design verification. Verification may include design reviews, alternate calculations, or qualification testing. Qualification tests are conducted in accordance with approved test programs or procedures.
Design interface controls are established and adequate.
Design, specification, and procedure changes are reviewed and approved in the same manner as the original issue. In a case where a proposed design change potentially affects licensed conditions, the Quality Assurance Program shall provide for ensuring that licensing considerations have been reviewed and are complied with or otherwise reconciled by amending the license.
Design errors and deficiencies are documented, corrected and corrective action to prevent recurrence is taken.
Design organization(s) and their responsibilities and authorities are delineated and controlled through written procedures.
3 DOE, Code of Federal Regulations, 10 CFR 830.122, Quality Assurance Criteria, U.S. Department of Energy, Washington, D.C., 2006.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                Rev. 16, May 2021 Materials, parts, equipment, and processes essential to the function of items that are important to safety are selected and reviewed for suitability of application.
Computer programs used for design analysis or verification are controlled in accordance with approved procedures. These procedures provide for verification of the accuracy of computer results and for the assessment and resolution of reported computer program errors.
9.4 Procurement Document Control As required by the INL Contractor Quality Program, procurement/acquisition processes and related document control activities are established and implemented to satisfy requirements of the QAPD. Requirements are to be in accordance with:
10 CFR 830.122(d), Criterion 4 - Management/Documents and Records 10 CFR 830.122(g), Criterion 7 - Performance/Procurement DOE Order 414C, CRD, Attachment 1, 2.a.(4), Criterion 4 - Documents and Records DOE Order 414C, CRD, Attachment 1, 2.b.(3), Criterion 7 - Procurement DOE Guide 414.1-3, Suspect/Counterfeit Items.
Processes and procedures are in place to ensure appropriate levels of quality are achieved in procurement of material, equipment, and services. Quality Level and Quality Category designations assigned by the Design Authority grade the application of QA requirements for procurements based on radiological material at risk, mission importance, safety of workers, public, environment, and equipment, and other differentiating criteria. Implementing procedures provide the logic process for determining Quality Levels used in procurement of equipment and subcontracting of services. Procedures ensure processes address document preparation and document control, and records management to meet regulatory requirements. Procurement records are kept in a manner that satisfies regulatory requirements.
INL Contractor procurement actions for packaging and spare parts shall be controlled. Contracts and Purchase Orders for packaging and spare parts shall require the selected vendor to implement and maintain an NRC approved 10CFR71, Subpart H QA Program.
Implementing procedures ensure procurement documents are prepared to clearly define applicable technical and quality assurance requirements including codes, standards, regulatory requirements and commitments, and contractual requirements. These documents serve as the principal documents for procurement of structures, systems and components, and related services for use in design, fabrication, maintenance and operation, inspection and testing of storage and/or transportation systems. Procedures ensure purchased material, components, equipment, and services adhere to applicable requirements. Furthermore:
The assignment of quality requirements through procurement documents is administered and controlled.
Procurement activities are performed in accordance with approved procedures delineating requirements for preparation, review, approval, and control of procurement documents.
Revisions to procurement documents are reviewed and approved by the same cognizant groups as the original document.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                    Rev. 16, May 2021 Quality requirements are included in quality-related purchase orders as applicable to the scope of the procurement referencing 10 CFR 71, Subpart H or other codes and standards, as appropriate.
INL Contractor procurement documents will require suppliers to convey appropriate quality assurance program requirements to sub-tier suppliers.
INL Contractor procurement documents will include provisions that suppliers either maintain or supply those QA records which provide evidence of conformance to the procurement documents. Additionally, procurement documents shall designate the supplier documents required for submittal to INL for review and/or approval.
INL shall maintain the right of access to supplier facilities and performance of source surveillance and/or audit activities, as applicable. A statement to this effect is to be included in procurement documents.
INL shall require the Supplier to warrant that all items furnished under the Contract are genuine (i.e., new, not refurbished, not counterfeit) and match the quality, test reports, markings and/or fitness for intended use as required by the Contract. Any materials furnished as part of the Contract which has been previously found to be suspect/counterfeit by the government or other duly recognized agency, shall not be used.
Procurement documents shall also address the applicability of the provisions of 10 CFR 21 for the Reporting of Defects and Noncompliances.
9.5 Instructions, Procedures, and Drawings As required by the INL Contractor Quality Program, instructions, procedures, and drawing work processes and applicable quality improvement activities shall be established and implemented to satisfy the requirements of the QAPD. These requirements are to be in accordance with:
10 CFR 830.122(c), Criterion 3 - Management/Quality Improvement 10 CFR 830.122(e), Criterion 5 - Performance/Work Processes DOE Order 414C, CRD, Attachment 1, 2.a.(3), Criterion 3 - Quality Improvement DOE Order 414C, CRD, Attachment 1, 2.b.(1), Criterion 5 - Work Processes.
Requirements are implemented to ensure processes and procedures are in place that achieve quality objectives and ensure appropriate levels of quality and safety are applied to critical components of packaging and transportation systems utilizing a graded approach. The program shall ensure processes and procedures in place to identify and correct problems associated with transportation and packaging activities.
Implementing procedures shall be established to ensure that methods for complying with each of the applicable criteria of 10 CFR 71, Subpart H, as applicable, for activities affecting quality during design, fabrication, inspection, testing, use and maintenance are specified in instructions, procedures, and/or drawings. In addition:
Instructions, procedures, and drawings shall be developed, reviewed, approved, utilized, and controlled in accordance with the requirements of approved procedures. These 9-13
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                Rev. 16, May 2021 instructions, procedures, and drawings shall include appropriate quantitative and qualitative acceptance criteria.
Changes to instructions, procedures and drawings, are developed, reviewed, approved, utilized and controlled using the same requirements and controls as applied to the original documents.
Compliance with these approved instructions, procedures and drawings is mandatory for INL personnel while performing activities affecting quality.
Specific activities by INL regarding preparation of packaging for use, repair, rework, maintenance, loading contents, unloading contents, and transport, must be accomplished in accordance with written and approved instructions, procedures, specifications, and/or drawings.
These documents must identify appropriate inspection and hold points and emphasize those characteristics that are important to safety and quality. Transportation package procedures are to be developed and reviewed by technical and quality staff and shall be approved by appropriate levels of management.
9.5.1 Preparation and Use Activities concerning loading and shipping are performed in accordance with written operating procedures developed by the user and approved by the package custodian. Packaging first-time usage tests, sequential loading and unloading operations, technical constraints, acceptance limits, and references are specified in the procedures. A pre-planned and documented inspection will be conducted to ensure that each loaded package is ready for delivery to the carrier.
9.5.2 Operating Procedure Changes Changes in operating procedures that affect the process must be approved at the same supervisory level as the initial issue.
9.5.3 Drawings Controlled drawings are shown in Appendix 1.3.2, Packaging General Arrangement Drawings, of this SAR. Implementation of design revisions is discussed in SAR Section 9.3, Package Design Control.
9.6 Document Control As required by the INL Contractor Quality Program, document control activities shall be established and implemented to satisfy the requirements of the QAPD. These requirements are to be in accordance with:
10 CFR 830.122(d), Criterion 4 - Management/Documents and Records DOE Order 414C, CRD, Attachment 1, 2.a.(4), Criterion 4 - Documents and Records.
Requirements are implemented to ensure processes and procedures are in place to address document, document control, and for the management of records. Records (engineering, test 9-14
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 reports, user instructions, etc.) must be maintained in a manner that conforms to regulatory requirements.
Document control activities related to the design, procurement, fabrication, and testing of ATR FFSC components; and SAR preparation shall be controlled.
Implementing procedures shall be established to control the issuance of documents that prescribe activities affecting quality and to assure adequate review, approval, release, distribution, use of documents and their revisions. Controlled documents may include, but are not limited to:
Design specifications Design and fabrication drawings Special process specifications and procedures QA Program Manuals/Plans, etc.
Implementing procedures Test procedures Operational test procedures and data.
Requirements shall ensure changes to documents, which prescribe activities affecting quality, are reviewed and approved by the same organization that performed the initial review and approval, or by qualified responsible organizations. Documents that prescribe activities affecting quality are to be reviewed and approved for technical adequacy and inclusion of appropriate quality requirements prior to approval and issuance. Measures are taken to ensure that only current documents are available at the locations where activities affecting quality are performed prior to commencing the work.
Package users are responsible for establishment, development, review, approval, distribution, revision, and retention of their documents. Documents requiring control, the level of control, and the personnel responsibilities and training requirements are to be identified.
Packaging documents to be controlled include as a minimum:
Operating procedures Maintenance procedures Inspection and test procedures Loading and unloading procedures Preparation for transport procedures Repair procedures Specifications Fabrication records Drawings of packaging and components SAR and occurring supplements.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 Revisions are handled in a like manner as the original issue. Only the latest revisions must be available for use.
Documentation received from the supplier for each package must be filed by package serial number. These documents are to be retained in the users facility.
9.7 Control Of Purchased Material, Equipment And Services As required by the INL Contractor Quality Program, the control of purchased material, equipment and services and applicable quality improvement activities shall be established and implemented to satisfy the requirements of the QAPD. These requirements are to be in accordance with:
10 CFR 830.122(c), Criterion 3 - Management/Quality Improvement 10 CFR 830.122(g), Criterion 7 - Performance/Procurement 10 CFR 830.122(h), Criterion 8 - Performance/Inspection and Acceptance Testing DOE Order 414C, CRD, Attachment 1, 2.b.(3), Criterion 3 - Quality Improvement DOE Order 414C, CRD, Attachment 1, 2.b.(3), Criterion 7 - Procurement DOE Order 414C, CRD, Attachment 1, 2.b.(4), Criterion 8 - Inspection and Acceptance Testing.
Requirements are implemented to ensure processes and procedures are in place to ensure appropriate inspections and tests are applied prior to acceptance or use of the packaging or component, and to identify the status of packaging items, components, etc. Requirements shall ensure processes and procedures are in place such that appropriate levels of quality are achieved in the procurement of material, equipment, and services. Quality Level and Quality Category designations by the Design Authority are used to grade the application of QA requirements of procurements based on radiological material at risk, mission importance, safety of workers, public, environment, and equipment, and other differentiating criteria. Requirements shall ensure processes and procedures in place to identify and correct problems associated with transportation and packaging activities.
Activities related to the control of purchased material, equipment and services shall be controlled. Control of purchased material, equipment, and services consist of the following elements:
Implementing procedures shall be established to assure that purchased material, equipment and services conform to procurement documents.
Procurement documents shall be reviewed and approved by authorized personnel for acceptability of proposed suppliers based on the quality requirements of the item/activity being purchased.
As required, audits and/or surveys are conducted to determine supplier acceptability.
These audits/surveys are based on one or all of the following criteria: the suppliers capability to comply with the requirements of 10 CFR 71, Subpart H that are applicable to the scope of work to be performed; a review of previous records to establish the past performance of the supplier; and/or a survey of the suppliers facilities and review of the 9-16
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 suppliers QA Program to assess adequacy and verify implementation of quality controls consistent with the requirements being invoked.
Qualified personnel shall conduct audits and surveys. Audit/survey results are to be documented and retained as Quality Assurance Records. Suppliers are re-audited and/or re-evaluated at planned intervals to verify that they continue to comply with quality requirements and to assess the continued effectiveness of their QA Program.
Additionally, interim periodic evaluations are to be performed of supplier quality activities to verify implementation of their QA Program.
Suppliers are required to provide objective evidence that items or services provided meet the requirements specified in procurement documents. Items are properly identified to appropriate records that are available to permit verification of conformance with procurement documents. Any procurement requirements not met by suppliers shall be reported to INL Contractor Quality Management for assessment of the condition. These conditions are reviewed by technical and quality personnel to assure that they have not compromised the quality or service of the item.
Periodic surveillance of supplier in-process activities is performed as necessary, to verify supplier compliance with the procurement documents. When deemed necessary, the need for surveillance is noted in approved quality or project planning documents.
Surveillances are to be performed and documented in accordance with approved procedures. Personnel performing surveillance of supplier activities are to be trained and qualified in accordance with approved procedures.
Quality planning for the performance of source surveillance, test, shipping and/or receiving inspection activities to verify compliance with approved design and licensing requirements, applicable 10 CFR 71 criteria, procurement document requirements, or contract specifications is to be performed in accordance with approved procedures.
For commercial off-the-shelf items, where specific quality controls appropriate for nuclear applications cannot be imposed in a practical manner, additional quality verification shall be performed to the extent necessary to verify the acceptability and conformance of an item to procurement document requirements. When dedication of a commercial grade item is required for use in a quality-related application, such dedication shall be performed in accordance with approved procedures.
To ensure compliance with procurement requirements, control measures shall include verification of supplier capability and verification of item or service quality. Procurements of ATR FFSC components are required to be placed with pre-qualified and selected vendors. The vendors QA Plan must address the requirements of 10 CFR 71, Subpart H and defined requirements. A graded approach is used based on the QA Levels established in Table 9.2-2.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 The approach used to control the procurement of items and services must include the following:
Source evaluation and selection Evaluation of objective evidence of quality furnished by the supplier Source inspection Audit Examination of items or services upon delivery or completion.
9.8 Identification And Control Of Material, Parts And Components As required by the INL Contractor Quality Program, activities concerning the identification and control of material, parts, and components shall be established and implemented to satisfy the requirements of QAPD. These requirements are to be in accordance with:
10 CFR 830.122(e), Criterion 5 - Performance/Work Processes 10 CFR 830.122(g), Criterion 7 - Performance/Procurement 10 CFR 830.122(h), Criterion 8 - Performance/Inspection and Acceptance Testing DOE Order 414C, CRD, Attachment 1, 2.b.(1), Criterion 5 - Work Processes DOE Order 414C, CRD, Attachment 1, 2.b.(3), Criterion 7 - Procurement DOE Order 414C, CRD, Attachment 1, 2.b.(4), Criterion 8 - Inspection and Acceptance Testing.
Requirements are implemented to ensure processes and procedures are in place that achieve quality objectives and ensure appropriate levels of quality and safety are applied to critical components of packaging and transportation systems utilizing a graded approach. The program also ensures processes and procedures are in place such that appropriate inspections and tests are applied prior to acceptance or use of the packaging or component, and to identify the status of packaging items, and components. The program shall ensure processes and procedures are in place to ensure appropriate levels of quality are achieved in the procurement of material, equipment, and services.
Activities related to the identification and control of material, parts and components shall be controlled. The requirements for identification and control of material, parts, and components consist of the following elements:
Implementing procedures are established to identify and control materials, parts, and components. These procedures assure identification of items by appropriate means during fabrication, installation, and use of the items and prevent the inadvertent use of incorrect or defective items.
Requirements for identification are established during the preparation of procedures and specifications.
Methods and location of identification are selected to not adversely affect the quality of the item(s) being identified.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 Items having limited shelf or operating life are controlled to prevent their inappropriate use.
Control and identification must be maintained either directly on the item or within documents traceable to the item to ensure that only correct and acceptable items are used. When physical identification is not practical, other appropriate means of control must be established such as bagging, physical separation, or procedural control. Each packaging unit shall be assigned a unique serial number after fabrication or purchase. All documentation associated with subsequent storage, use, maintenance, inspection, acceptance, etc., must refer to the assigned serial number. Verification of acceptance status is required prior to use. Items that are not acceptable must be controlled accordingly. Control of nonconforming items is addressed in Section 9.15, Nonconforming Parts, Materials, or Components.
Each ATR FFSC package will be conspicuously and durably marked with information identifying the package owner, model number, unique serial number, and package gross weight, in accordance with 10 CFR 71.85(c).
Replacement parts must be identified to ensure correct application. Minute items must be individually packaged with the package marked with the part identification and traceability information.
9.9 Control Of Special Processes As required by the INL Contractor Quality Program, activities for the control of special processes shall be established and implemented to satisfy the requirements of the QAPD. These requirements are to be in accordance with:
10 CRF 830.122(b), Criterion 2 - Management/Personnel Training and Qualifications 10 CFR 830.122(e), Criterion 5 - Performance/Work Processes 10 CFR 830.122(g), Criterion 7 - Performance/Procurement DOE Order 414C, CRD, Attachment 1, 2.a.(2), Criterion 2 - Personnel Training and Qualifications DOE Order 414C, CRD, Attachment 1, 2.b.(1), Criterion 5 - Work Processes DOE Order 414C, CRD, Attachment 1, 2.b.(3), Criterion 7 - Procurement.
Requirements will be implemented to ensure only trained and qualified personnel perform transportation and packaging activities. The program shall ensure processes and procedures are in place that achieve quality objectives and ensure appropriate levels of quality and safety are applied to critical components of packaging and transportation systems utilizing a graded approach.
Activities related to the control of special processes shall be controlled. The requirements for control of special processes consist of the following elements:
Implementing procedures shall be established to control special processes used in the fabrication and inspection of storage/transport systems. These processes may include welding, non-destructive examination, or other special processes as identified in procurement documents.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                        Rev. 16, May 2021 Special processes are performed in accordance with approved procedures.
Personnel who perform special processes shall be trained and qualified in accordance with applicable codes, standards, specifications, and/or other special requirements.
Records of qualified procedures and personnel are to be maintained and kept current by the organization that performs the special processes.
Package users are responsible to ensure special processes for welding and nondestructive examination of the ATR FFSC during fabrication, use, and maintenance are controlled.
Equipment used in conduct of special processes must be qualified in accordance with applicable codes, standards, and specifications. Special process operations must be performed by qualified personnel and accomplished in accordance with written process sheets or procedures with recorded evidence of verification when applicable. Qualification records of special process procedures, equipment, and personnel must be maintained.
Welders, weld procedures, and examination personnel are to be qualified in accordance with the appropriate articles of ASME BPVC, Section IX, Welding and Brazing Qualifications;4 and ASME BPVC, Section V, Nondestructive Examination.5 Special processes for QA Level A and B items must be performed by qualified personnel in accordance with documented and approved procedures. Applicable special processes performed by an outside supplier such as welding, plating, anodizing, and heat treating, which are controlled by the suppliers quality program, are reviewed and/or witnessed in accordance with procurement requirements.
9.10 Internal Inspection As required by the INL Contractor Quality Program, internal inspection activities shall be established to satisfy the requirements of the QAPD. These requirements are to be in accordance with:
10 CRF 830.122(b), Criterion 2 - Management/Personnel Training and Qualifications 10 CFR 830.122(h), Criterion 8 - Performance/Inspection and Acceptance Testing DOE Order 414C, CRD, Attachment 1, 2.a.(2), Criterion 2 - Personnel Training and Qualifications DOE Order 414C, CRD, Attachment 1, 2.b.(4), Criterion 8 - Inspection and Acceptance Testing.
Requirements are implemented to ensure only trained and qualified personnel perform transportation and packaging activities. The program shall ensure processes and procedures are in place to ensure appropriate inspections and tests are applied prior to acceptance or use of the packaging or component, and to identify the status of packaging items, components, etc.
4 ASME, 2004, American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section IX, Welding and Brazing Qualifications, American Society of Mechanical Engineers, New York, NY 5
ASME, 2004, American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section V, Nondestructive Examination, American Society of Mechanical Engineers, New York, NY 9-20
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 Activities related to internal inspection shall be controlled. The program requirements for control of internal inspection consist of the following elements:
Implementing procedures shall be established to assure that inspection or surveillance is performed to verify that materials, parts, processes, or other activities affecting quality conform to documented instructions, procedures, specifications, drawings, and/or procurement documents.
Personnel performing inspection and surveillance activities shall be trained and qualified in accordance with written approved procedures.
Inspections and surveillances are to be performed by individuals other than those who performed or supervised the subject activities.
Inspection or surveillance and process monitoring are both required where either one, by itself, will not provide assurance of quality.
Modifications and/or repairs to and replacements of safety-related and important-to-safety structures, systems, and components are inspected in accordance with the original design and inspection requirements or acceptable alternatives.
Mandatory hold points, inspection equipment requirements, acceptance criteria, personnel qualification requirements, performance characteristics, variable and/or attribute recording instructions, reference documents, and other requirements are considered and included, as applicable, during inspection and surveillance planning.
9.10.1      Inspections During Fabrication Specific inspection criteria are incorporated into the drawings for the ATR FFSC packaging.
Inspection requirements for fabrication are divided into two responsible areas that document that an accepted ATR FFSC package conforms to tested and certified design criteria. These two areas are:
In-process inspections performed by the fabricator.
Independent surveillance of fabrication activities performed by individuals acting on behalf of the purchaser.
The vendor (fabricator) is required to submit Manufacturing/Fabrication Plans prior to the start of fabrication for approval by the customer. These plans shall be used as a tool for establishing witness and hold points. A review for compliance with procurement documents is normally performed as part of the surveillance function at the vendors facility. The plans shall define how fabrications and inspections are to be performed, processes to be engaged. Inspections must be documented and records delivered in individual data packages accompanying the package in accordance with the procurement specification.
Independent surveillance activities will be performed by qualified personnel selected with approval of the customer.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 9.10.2      Inspections During Initial Acceptance and During Service Life Independent inspections are performed upon receipt of the ATR FFSC packaging prior to first usage (implemented by package user procedures) and on an annual basis. Post-loading inspections are also performed prior to shipment. Inspection to be implemented by the package user (by qualified independent inspection personnel) must include the following:
Acceptance - Ensure compliance with procurement documents. Per Chapter 8, Acceptance Tests and Maintenance Program of this SAR, perform (as applicable) first-time-usage inspections, and weld examinations.
Operation - Verify proper assembly and verify that post-load leak testing (if applicable) is carried out as discussed in Chapter 7, Package Operations, of this SAR.
Maintenance - Ensure adequate packaging maintenance to ensure that performance is not impaired as discussed in Chapter 8, Acceptance Tests and Maintenance Program of this SAR.
Final - Verify proper contents, assembly, marking, shipping papers, and implementation of any special instructions.
9.11 Test Control As required by the INL Contractor Quality Program, test control activities shall be established and implemented to satisfy the requirements of the QAPD. These requirements are to be in accordance with:
10 CFR 830.122(e), Criterion 5 - Performance/Work Processes DOE Order 414C, CRD, Attachment 1, 2.b.(1), Criterion 5 - Work Processes.
Requirements are implemented to ensure processes and procedures are in place that achieve quality objectives and ensure appropriate levels of quality and safety are applied to critical components of packaging and transportation systems utilizing a graded approach.
Activities related to test control shall be controlled. The requirements for test control consist of the following elements:
Implementing procedures shall be established to assure that required proof, acceptance, and operational tests, as identified in design or procurement documents, are performed and appropriately controlled.
Test personnel shall have appropriate training and shall be qualified for the level of testing which they are performing. Personnel shall be qualified in accordance with approved, written instructions, procedures, and/or checklists.
Tests are performed by qualified personnel in accordance with approved, written instructions, procedures, and/or checklists. Test procedures are to contain or reference the following information, as applicable:
            - Acceptance criteria contained in the applicable test specifications, or design and procurement documents.
            - Instructions for performance of tests, including environmental conditions.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021
            - Test prerequisites such as test equipment, instrumentation requirements, personnel qualification requirements, fabrication, or operational status of the items to be tested.
            - Provisions for data recording and records retention.
Test results are to be documented and evaluated to ensure that acceptance criteria have been satisfied.
Tests to be conducted after modifications, repairs, or replacements of safety-related and important-to-safety structures, systems, or components are to be performed in accordance with the original design and testing requirements or acceptable alternatives.
Tests are required when it is necessary to demonstrate that an item or process will perform satisfactorily. Test procedures must specify the objectives of the tests, testing methods, required documentation, and acceptance criteria. Tests to be conducted by vendors at vendor facilities must be specified in procurement documents. Personnel conducting tests, test equipment, and procedures must be qualified and records attesting to qualification retained.
9.11.1      Acceptance and Periodic Tests The fabricator must supply QA documentation for the fabrication of each ATR FFSC packaging in accordance with applicable drawings, specifications, and/or other written requirements.
The package user must ensure required ATR FFSC packaging inspections and tests are performed prior to first usage.
Periodic testing, as applicable, will be performed to ensure the ATR FFSC packaging performance has not deteriorated with time and usage. The requirements for the periodic tests are given in the Chapter 8, Acceptance Tests and Maintenance Program of this SAR. The results of these tests are required to be documented and maintained with the specific packaging records by the package user.
9.11.2      Packaging Nonconformance Packaging that does not meet the inspection criteria shall be marked or tagged as nonconforming, isolated, and documented in accordance with Section 9.15, Nonconforming Parts, Materials, or Components. The packaging must not be used for shipment until the nonconformance report has been properly dispositioned in accordance with Section 9.15.
9.12 Control Of Measuring And Test Equipment As required by the INL Contractor Quality Program, activities pertaining to the control of measuring and test equipment shall be established and implemented to satisfy the requirements of the QAPD. These requirements are to be in accordance with:
10 CFR 830.122(h), Criterion 8 - Performance/Inspection and Acceptance Testing DOE Order 414C, CRD, Attachment 1, 2.b.(4), Criterion 8 - Inspection and Acceptance Testing.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 Requirements are implemented to ensure processes and procedures are in place to ensure appropriate inspections and tests are applied prior to acceptance or use of the packaging or component, and to identify the status of packaging items, components, etc.
Activities pertaining to the control of measuring and test equipment shall be controlled. The requirements for control of measuring and test equipment shall consist of the following elements:
Implementing procedures shall be established to assure that tools, gages, instruments and other measuring and testing devices (M&TE) used in activities affecting quality are properly controlled, calibrated and adjusted to maintain accuracy within required limits.
M&TE are calibrated at scheduled intervals against certified standards having known valid relationships to national standards. If no national standards exist, the basis for calibration shall be documented. Calibration intervals are based on required accuracy, precision, purpose, amount of use, stability characteristics and other conditions that could affect the measurements.
Calibrations are to be performed in accordance with approved written procedures.
Inspection, measuring and test equipment are to be marked to indicate calibration status.
M&TE are to be identified, labeled or tagged indicating the next required calibration due date, and traceable to calibration records.
If M&TE is found to be out of calibration, an evaluation shall be performed and documented regarding the validity of inspections or tests performed and the acceptability of items inspected or tested since the previous acceptable calibration. The current status of M&TE is to be recorded and maintained. Any M&TE that is consistently found to be out of calibration shall be repaired or replaced.
Special calibration and control measures on rules, tape measures, levels and other such devices are not required where normal commercial practices provide adequate accuracy.
9.13 Handling, Storage, And Shipping Control As required by the INL Contractor Quality Program, handling, storage, and shipping control activities shall be established and implemented to satisfy the requirements of the QAPD. These requirements are to be in accordance with:
10 CFR 830.122(e), Criterion 5 - Performance/Work Processes DOE Order 414C, CRD, Attachment 1, 2.b.(1), Criterion 5 - Work Processes.
Requirements are implemented to ensure processes and procedures are in place that achieve quality objectives and ensure appropriate levels of quality and safety are applied to critical components of packaging and transportation systems utilizing a graded approach.
Activities pertaining to handling, storage, and shipping shall be controlled. The requirements for handling, storage, and shipping control consist of the following elements:
Implementing procedures shall be established to assure that materials, parts, assemblies, spare parts, special tools, and equipment are handled, stored, packaged, and shipped in a manner to prevent damage, loss, loss of identity, or deterioration.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                    Rev. 16, May 2021 When necessary, storage procedures address special requirements for environmental protection such as inert gas atmospheres, moisture control, temperature levels, etc.
Package users shall ensure that components associated with the ATR FFSC are controlled to prevent damage or loss, protected against damage or deterioration, and provide adequate safety of personnel involved in handling, storage, and shipment (outgoing and incoming) operations.
Handling, storage, and shipping must be accomplished in accordance with written and approved instructions, procedures, specifications, and/or drawings. These documents must identify appropriate information regarding shelf life, environment, temperature, cleaning, handling, and preservation, as applicable, to meet design, regulatory, and/or DOE shipping requirements.
Preparation for loading, handling, and shipment will be done accordance with approved procedures to ensure that all requirements have been met prior to delivery to a carrier. A package ready for shipment must conform to its shipping paper.
Empty packages, following usage, must be checked and decontaminated if required. Each package must be inspected, reconditioned, or repaired, as appropriate, in accordance with approved written procedures before storing or loading. Empty ATR FFSC packagings are to be tagged with EMPTY labels and stored in designated protected areas in order to minimize environmental effects on the containers.
Routine maintenance on the ATR FFSC packaging may be performed as deemed necessary by package users and is limited to cleaning, rust removal, painting, light metal working to restore the original contours and replacement of damaged, worn, or malfunctioning components. Spare components will be placed in segregated storage to maintain proper identification and to avoid misuse.
9.14 Inspection, Test, And Operating Status As required by the INL Contractor Quality Program, inspection, test, and operating status activities shall be established and implemented to satisfy the requirements of the QAPD. These requirements are to be in accordance with:
10 CFR 830.122(e), Criterion 5 - Performance/Work Processes 10 CFR 830.122(h), Criterion 8 - Performance/Inspection and Acceptance Testing DOE Order 414C, CRD, Attachment 1, 2.b.(1), Criterion 5 - Work Processes DOE Order 414C, CRD, Attachment 1, 2.b.(4), Criterion 8 - Inspection and Acceptance Testing.
Requirements are implemented to ensure processes and procedures are in place that achieve quality objectives and ensure appropriate levels of quality and safety are applied to critical components of packaging and transportation systems utilizing a graded approach. In addition, processes and procedures shall be in place to ensure appropriate inspections and tests are applied prior to acceptance or use of the packaging or component, and to identify the status of packaging items, components, etc.
Activities pertaining to inspection, test, and operating status activities shall be controlled. The requirements for inspection, test, and operating status consist of the following elements:
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 Implementing procedures shall be established to assure that the inspection and test status of materials, items, structures, systems, and components throughout fabrication, installation, operation, and test are clearly indicated by suitable means, (e.g., tags, labels, cards, form sheets, check lists, etc.).
Bypassing of required inspections, tests, or other critical operations is prevented through the use of approved instructions or procedures As appropriate, the operating status of nonconforming, inoperative or malfunctioning components of a storage/transport system is indicated to prevent inadvertent operation.
The application and removal of status indicators is performed in accordance with approved instructions and procedures.
Any nonconforming items are identified and controlled in accordance with Section 9.15, Nonconforming Parts, Materials, or Components, of this SAR.
Package users shall ensure that the status of inspection and test activities are identified on the item or in documents traceable to the item to ensure that proper inspections or tests have been performed and that those items that do not pass inspection are not used. The status of fabrication, inspection, test, assembly, and refurbishment activities must be identified in documents traceable to the package components.
Measures established in specifications, procedures, and other instructions shall ensure that the following objectives are met:
QA personnel responsible for oversight of packaging inspections can readily ascertain the status of inspections, tests, and/or operating conditions.
No controlled items are overlooked.
Inadvertent use or installation of unqualified items is prevented.
Documentation is complete.
9.15 Nonconforming Materials, Parts, or Components As required by the INL Contractor Quality Program, control of nonconforming materials, parts, or components shall be established and implemented to satisfy the requirements of the QAPD.
These requirements are to be in accordance with:
10 CFR 830.122(c), Criterion 3 - Management/Quality Improvement DOE Order 414C, CRD, Attachment 1, 2.b.(3), Criterion 3 - Quality Improvement.
Requirements are implemented to ensure that processes and procedures are in place to identify and correct problems associated with transportation and packaging activities.
Activities pertaining to the control of nonconforming materials, parts, or components shall be controlled. The requirements for nonconforming materials, parts, or components consist of the following elements:
Implementing procedures shall be established to control materials, parts, and components that do not conform to requirements to prevent their inadvertent use during fabrication or during service.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                Rev. 16, May 2021 Nonconforming items include those items that do not meet specification or drawing requirements. Additionally, nonconforming items include items not fabricated or tested (1) in accordance with approved written procedures, (2) by qualified processes, or (3) by qualified personnel; where use of such procedures, processes, or personnel is required by the fabrication, test, inspection, or quality assurance requirements.
Nonconforming items are identified and/or segregated to prevent their inadvertent use until properly dispositioned. The identification of nonconforming items is by marking, tagging, or other methods that do not adversely affect the end use of the item. The identification shall be legible and easily recognizable. When identification of each nonconforming item is not practical, the container, package, or segregated storage area, as appropriate, is identified.
Nonconforming conditions are documented in NCRs and affected organizations are to be notified. The nonconformance report shall include a description of the nonconforming condition. Nonconforming items are dispositioned as use-as-is, reject, repair, or rework.
Inspection or surveillance requirements for nonconforming items following rework, repair are detailed in the nonconformance reports and approved following completion of the disposition.
Acceptability of rework or repair of nonconforming materials, parts, and components is verified by re-inspecting and/or re-testing the item to the original requirements or equivalent inspection/testing methods. Inspection, testing, rework, and repair methods are to be documented and controlled.
The disposition of nonconforming items as use-as-is or repair shall include technical justification and independent verification to assure compliance with design, regulatory, and contractual requirements.
Items dispositioned as rework or repair are reinspected and retested in accordance with the original inspection and test requirements or acceptable alternatives that comply with the specified acceptance criteria.
When specified by contract requirements, nonconformances that result in a violation of client contract or specification requirements shall be submitted for client approval.
Nonconformance reports are made part of the inspection records and are periodically reviewed to identify quality trends. Unsatisfactory quality trends are documented on a Corrective Action Report (CAR) as detailed in Section 9.16, Corrective Action, of this SAR. The results of these reviews are to be reported to management.
Nonconformance reports relating to internal activities are issued to management of the affected organization. The appropriate Quality Assurance Manager shall approve the disposition and performs follow-up activities to assure proper closure.
Compliance with the evaluation and reporting requirements of 10 CFR 21 related to defects and noncompliances are to be controlled by approved procedures.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 9.16 Corrective Action As required by the INL Contractor Quality Program, requirements for corrective action shall be established and implemented to satisfy the requirements of the QAPD. These requirements are to be in accordance with:
10 CFR 830.122(c), Criterion 3 - Management/Quality Improvement DOE Order 414C, CRD, Attachment 1, 2.b.(3), Criterion 3 - Quality Improvement.
Requirements are implemented to ensure that processes and procedures are in place to identify and correct problems associated with transportation and packaging activities.
Activities pertaining to corrective actions shall be controlled. The requirements for corrective action consist of the following elements:
Implementing procedures shall be established to identify significant conditions adverse to quality. Significant and/or repetitive failures, malfunctions and deficiencies in material, components, equipment, and operations are to be promptly identified and documented on a Corrective Action Reports (CARs) and reported to appropriate management. The cause of the condition and corrective action necessary to prevent recurrence are identified, implemented, and followed up to verify corrective action is complete and effective.
The INL Contractor Quality Assurance Director (DQA) is responsible for ensuring implementation of the corrective action program, including follow up and closeout actions. The DQA may delegate certain activities in the Corrective Action process to others.
9.17 Quality Assurance Records As required by the INL Contractor Quality Program, activities associated with QA records shall be established and implemented to satisfy the requirements of the QAPD. These requirements are to be in accordance with:
10 CRF 830.122(b), Criterion 2 - Management/Personnel Training and Qualifications 10 CFR 830.122(d), Criterion 4 - Management/Documents and Records 10 CFR 830.122(e), Criterion 5 - Performance/Work Processes 10 CFR 830.122(h), Criterion 8 - Performance/Inspection and Acceptance Testing DOE Order 414C, CRD, Attachment 1, 2.a.(2), Criterion 2 - Personnel Training and Qualifications DOE Order 414C, CRD, Attachment 1, 2.a.(4), Criterion 4 - Documents and Records DOE Order 414C, CRD, Attachment 1, 2.b.(1), Criterion 5 - Work Processes DOE Order 414C, CRD, Attachment 1, 2.b.(4), Criterion 8 - Inspection and Acceptance Testing.
Requirements are implemented to ensure that only trained and qualified personnel perform transportation and packaging activities. The program shall ensure processes and procedures are 9-28
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                    Rev. 16, May 2021 in place to address document preparation, document control, and management of records. In addition, the program ensures processes and procedures are in place which achieves quality objectives and appropriate levels of quality and safety are applied to critical components of packaging and transportation systems utilizing a graded approach. Finally, the program ensures processes and procedures are in place to identify appropriate inspections and tests are applied prior to acceptance or use of the package or component, and to identify the status of packaging items, components, etc.
Quality assurance records shall be controlled. The requirements for quality assurance records consist of the following elements:
Implementing procedures shall be established to assure control of quality records. The purpose of the Quality Assurance Records system is to assure that documented evidence relative to quality related activities is maintained and available for use by INL Contractor, its customers, and/or regulatory agencies, as applicable.
Approved procedures identify the types of documents to be retained as QA records, as well as those to be retained by the originating organization. Lifetime and Non-Permanent records are retained by Records Management (RMA) or its customers, as appropriate.
Records are identified, indexed, and stored in accessible locations.
QA Records are maintained for periods specified to furnish evidence of activities affecting the quality of structures, systems, and components that are safety-related or important-to-safety. These records include records of design, procurement, fabrication, assembly, inspection, and testing.
Maintenance records shall include the use of operating logs; results of reviews, inspections, tests, and audits; results from monitoring of work performance and material analyses; results of maintenance, modification, and repair activities; qualification of personnel, procedures, and equipment; records of calibration of measuring and test equipment; and related instructions, procedures, and drawings.
Requirements for indexing, record retention period, storage method(s) and location(s),
classification, preservation measures, disposition of nonpermanent records, and responsibility for safekeeping are specified in approved procedures. Record storage facilities are established to prevent destruction of records by fire, flood, theft, and deterioration due to environmental conditions (such as temperature, humidity, or vermin).
As an alternative, two identical sets of records (dual storage) may be maintained at separate locations.
INL shall retain required records for at least three (3) years beyond the date of last engagement of activities.
9.17.1      General Sufficient records must be maintained by package users to furnish evidence of quality of items and of activities affecting quality. QA records that must be retained for the lifetime of the packaging include:
Appropriate production-related records that are generated throughout the package manufacturing and fabrication process 9-29
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 Records demonstrating evidence of operational capability; e.g., completed acceptance tests and inspections Records verifying repair, rework, and replacement Audit reports, and corrective actions Records that are used as a baseline for maintenance Records showing evidence of delivery of packages to a carrier and proof that all DOT requirements were satisfied.
9.17.2      Generating Records Package user documents designated as QA records must be:
Legible Completed to reflect the work accomplished and relevant results or conclusions Signed and dated or otherwise authenticated by authorized personnel.
QA records should be placed in a records storage area as soon as is feasible to avoid loss or damage. Individual package QA records must be generated and maintained for each package by the package serial number.
9.17.3      Receipt, Retrieval, and Disposition of Records The RMA has overall responsibility for records management for the ATR FFSC. Package users are responsible for maintaining records while they are in process and for providing completed records to the RMA. A receipt control system shall be established, and records maintained in-house or at other locations are to be identifiable and retrievable and not disposed of until prescribed conditions are satisfied.
Records are to be available for inspection upon request.
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Docket No. 71-9330 ATR FFSC Safety Analysis Report                                    Rev. 16, May 2021 Table 9.17 Quality Assurance Records Retention Quality Assurance Record period Design and Fabrication Drawings                                  LOP+
Test Reports                                                    LOP+
Independent Design Review Comments                              LOP+
Safety Analysis Report for Packaging                            LOP+
Vendor Manufacturing and Inspection Plans                        LOP+
Material Test Report of Certification of Materials              LOP+
Welding Specifications and Procedures                            LOP+
Weld Procedure Qualification Record                              LOP+
Welder or Welding Operator Qualification Tests                  LOP+
Record of Qualification of Personnel Performing                  LOP+
Radiographic and PT Reports Weld Radiographs                                                LOP+
Liquid Penetrant Reports                                        LOP+
Dimensional Inspection Report for All Features                  LOP+
Visual and Dimensional Inspection upon Receipt of                LOP+
Packaging Package Loading Procedure                                          S+
Unloading Procedure                                                S+
Maintenance Procedures                                          LOP+
Repair Procedures                                                LOP+
Procurement Specifications                                      LOP+
Personnel Training and Qualification Documentation              LOP+
Maintenance Log                                                  LOP+
Corrective Action Reports                                        LOP+
Nonconformance Reports (and resolutions)                        LOP+
Incident Reports per 10 CFR 71.95                                LOP+
Preliminary Determinations per 10 CFR 71.85                        S+
Routine Determinations per 10 CFR 71.87                            S+
Shipment Records per 10 CFR 71.91(a), (b), (c), (d)                S+
LOP+ Lifetime of packaging plus 3 years        S+ Shipping date plus 3 years 9-31
 
Docket No. 71-9330 ATR FFSC Safety Analysis Report                                                  Rev. 16, May 2021 9.18 Audits As required by the INL Contractor Quality Program, audit requirements shall be established and implemented to satisfy the requirements of the QAPD. These requirements are to be in accordance with:
10 CFR 830.122(i), Criterion 9 - Assessment/Management Assessment 10 CFR 830.122(j), Criterion 10 - Assessment/Independent Assessment DOE Order 414C, CRD, Attachment 1, 2.c.(1), Criterion 9 - Management Assessment DOE Order 414C, CRD, Attachment 1, 2.c.(2), Criterion 10 - Independent Assessment.
Requirements are implemented to ensure management assessments are performed on a regular basis. Management assessments are planned and conducted in accordance with written procedures. In addition, the program will be independently assessed periodically in accordance with procedures.
Activities pertaining to audits and assessments shall be controlled. The requirements for audits and assessments consist of the following elements:
Implementing procedures shall be established to assure that periodic audits verify compliance with all aspects of the Quality Assurance Program and determine its effectiveness. Areas and activities to be audited, such as design, procurement, fabrication, inspection, and testing of storage/transportation systems, are to be identified as part of audit planning.
INL audits supplier Quality Assurance Programs, procedures, and implementation activities to evaluate and verify that procedures and activities are adequate and comply with applicable requirements.
Audits are planned and scheduled in a manner to provide coverage and coordination with ongoing Quality Assurance Program activities commensurate with the status and importance of the activities.
Audits are performed by trained and qualified personnel not having direct responsibilities in the areas being audited and are conducted in accordance with written plans and checklists. Audit results are documented and reviewed by management having responsibility for the area audited. Corrective actions and schedules for implementation are established and recorded. Audit reports include an objective evaluation of the quality-related practices, procedures, and instructions for the areas or activities being audited and the effectiveness of implementation.
Responsible management shall undertake corrective actions as a follow-up to audit reports when appropriate. The Quality Assurance Management (QAM) shall evaluate audit results for indications of adverse trends that could affect quality. When results of such assessments so indicate, appropriate corrective action will be implemented.
The QAM shall follow up on audit findings to assure that appropriate corrective actions have been implemented and directs the performance of re-audits when deemed necessary.
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Revision as of 15:54, 8 September 2021

Safety Analysis Report for the ATR-FFSC Package, Revision 16, May 2021, Part 3
ML21118B036
Person / Time
Site: 07109330
Issue date: 05/01/2021
From: Pierre Saverot
Storage and Transportation Licensing Branch
To:
PSaverot NMSS/DFM/STL 301.415.7505
Shared Package
ML21118B033 List:
References
Download: ML21118B036 (356)
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