ML18127B742
| ML18127B742 | |
| Person / Time | |
|---|---|
| Site: | 07109215 |
| Issue date: | 05/07/2018 |
| From: | Neutron Products |
| To: | Office of Nuclear Material Safety and Safeguards |
| Shared Package | |
| ML18127B734 | List: |
| References | |
| Download: ML18127B742 (22) | |
Text
5. SHIELDING EVALUATION 5.1 Description of Shielding Design 5.1.1 Design Features The principal shielding is provided by the shield/transfer container (S/TC} portion of the package. The shielding can be considered comprising three parts: the source holder\\ the drum into which the source holder fits, and the lead filled, S/TC Shell Assembly and End Covers.
The source(s) can be positioned either horizontally or vertically within the source chamber.
The holder or drawer is fabricated of one or more of the following shielding materials:
steel, depleted uranium, tungsten, lead, or brass. The remaining space in the Drum Assembly chambers is filled with shielding fabricated of steel, tungsten alloy, and/or lead of a dimension to provide the specified clearance tolerance during shipment. Thus, the Drum Assembly chambers are filled with shielding that is an inherent part of the source.capsule or the shipping packaging. As expected, and verified in Table 5.1.1, the dose rate from the package depends upon the specific shielding arrangement and source orientation, as well as the total activity.
The drum in which the source holder or drawer is carried is the second shielding barrier. The drum is an 8-3/16 inch diameter cylinder, 21-5/8 inches long, penetrated by tubes of various configurations which form the source chambers:
Drawing 240122 item 5 drum has three round through holes, each 2.56 inches in inside diameter. This configuration is authorized for 15,000 Ci of cobalt-60 or 20,600 Ci of cesium-137; Drawing 240122 item 4 drum has three larger round through holes, each 2.81 inches in inside diameter. This configuration is authorized for 9,500 Ci of cobalt-60 or 20,600 Ci of cesium-137; Drawing 240122 item 2 drum has two square through holes, each 3.09 inches on a side. This configuration is authorized for 6,300 Ci of cobalt-60 or 20,600 Ci of cesium-137.
1 In this section, the term source holder refers to either a traditional source drawer, in which the source is fixed in position within a full length, full diameter assembly which also provides axial shielding, or any other arrangement in which the source is kept in a shielded position by shielding/shield plugs which may or may not be physically attached to the source itself.
USA/9215/B(U) Consolidated Application for Renewal April 2018 Revision 6 Page 5-1
In all cases, the tubes are parallel to the axis of the drum and extend through its entire length. The axes of the round tubes are equally spaced circumferentially on a bolt circle of 1-3/4 inch radius concentric with the axis of the drum, while the square tubes are adjoining, with the common side centered in the drum. The source holders are slip fits into the source chambers. Frequently, only one source is carried per container. The source chambers not containing sources are loaded with shield plugs of steel, lead filled steel or tungsten alloy.
The drum fits into the 5/TC Shell Assembly which, along with the two End Covers, provides the third shielding barrier. Both the 5/TC Shell Assembly and End Covers are lead filled.
The bolted End Covers hold the drum tightly in place during transport.
The overpacl<, into which the 5/TC fits, contributes to dose reduction, principally by the geometric factor. The six-inch thickness of wood and 0.1 inch thickness of steel contribute little to the gamma shielding.
USA/9215/B{U) Consolidated Application for Renewal April 2018 Revision 6 Page 5-2
5.1.2 Summary Table of Maximum Radiation Levels The package can be loaded with many different source and shielding configurations, depending upon the particular application. As such, the maximum radiation levels for non-exclusive use shipments are the regulatory limits of 2 mSv/hr (200 mR/hr) at contact with the package, and a transport index of 10.
Normal Conditions of Package Surface 1 Meter from Package Surface Transport mSv/h (mrem/h) mSv/h (mrem/h)
Radiation Top Side Bottom Top Side Bottom 0.1 0.3 0.02 0.03 Gamma (lO}E 1.04 (104)
(30)E (2)E 0.14 {14)
{3)E Neutron n/a n/a n/a n/a n/a n/a Total 0.1 0.3 0.02 0.03 (10)E 1.04 (104)
(30)E (2)E 0.14 (14)
(3)E i
10 CFR71.47(a) limit I
2(200) 2(200) 2(200) 0.1 {10) 0.1 (10) 0.1 {10)
For transport index over 10, must be shipped exclusive use.
"E" denotes estimate. Dose rate in highest area {side} calculated. See 5.4 USA/9215/B(U) Consolidated Application for Renewal April 2018 Revision 6 Page 5-3
The change in shielding effectiveness under hypothetical accident conditions is due to shifting of the S/TC with reference to the outer surface of the package as a consequence of the 30-foot drop, puncture and fire. With a maximum estimated inner container shift of seven inches, the surface dose increases by a factor of 2 and the dose at 1 meter by about 25 percent. Both values are below the 10 CFR 71.51(a)(2) limit of 1,000 mR/hr at 1 meter for the hypothetical accident condition. There is no opportunity for any measurable shift of source or shielding within the S/TC itself under the most severe free drop condition.
Greater detail is provided in 5.4.4.
1 Meter from Package Surface Hypothetical Accident mSv/h (mrem/h)
Conditions Radiation Top Side Bottom Gamma 0.013 (1.3) E 0.18 (18) 0.038 {3.8) E Neutron n/a n/a n/a Total 0.013 (1.3} E 0.18 (18) 0.038 (3.8) 10 CFR71.51(a)(2) limit 10 (1000) 10 (1000) 10(1000)
"E" denotes estimate.
Dose rates at 1 m expected to increase by approximately 25% following HAC. See 5.4.4 5.2 Source Specification The only sources considered for radiation shielding design and evaluation are cobalt-60 and cesium-137.
5.2.1 Gamma Source The cesium-137 decay results in a single 661 keV photon. Each cobalt-GO disintegration produces two photons, having energies of 1.17 MeV and 1.33 MeV, for a total of 2.5 MeV per disintegration.
5.2.2 Neutron Source There is no neutron source transported in this package.
USA/9215/B(U) Consolidated Application for Renewal April 2018 Revision 6 Page 5-4
5.3 Shielding Model The shielding evaluation is based both on analytical models and on measurements made on the packages and determining the changes in dose rate due to the changes in geometry and materials.
The analysis employs a simple exponential attenuation model postulating an isotropic source.
Buildup factors were obtained from the Radiological Health Handbook, Revised Edition (January 1970). Values of material densities and mass attenuation coefficients are shown in Table 5.3.1 and were obtained from the same source. Streaming was also considered. The calculational models employed are described in Section 5.4. Subsequentfy a radiation survey of a referenced shipping package provided results which compared favorably with the calculations.
5.3.1 Configuration of Source and Shielding Figure 5.3.1 depicts an 5/TC with a source loading configuration. The areas of highest contact dose rate with the S/TC are generally around the steel flanges to which the End Covers are bolted, because a greater percentage of the shielding in these areas is provided by components fabricated out of steel as opposed to lead (for the rest of the 5/TC) and/or tungsten alloy (used in some source shields). This effect increases as the length of the source being transported increases.
As a result, the areas of highest contact dose rate with the package itself can generally be found on the overpack in the areas extending outward from the End Cover flanges.
As described in section 5.4, under hypothetical accident conditions, the contact dose rate with the package could increase by a factor of two when compared to contact dose rates under normal conditions due geometric considerations as the S/TC could shift closer to the outside edge of the package following the crushing of the WPJ rings due to the puncture and/or 30 foot drop and the charring effect of the fire.
USA/9215/B(U) Consolidated Application for Renewal April 2018 Revision 6 Page 5-5
SOURCE HOLDER SHELL LINER DRUM CASING -------
COVER COVER BOLTS BASE.
LIFTING LUGS ARE.A OF
~
I G \\-\\EST Dos£ RATE.
S/TC INNER CONTAINER STRUCTURAL COMPONENTS FIGURE 5.3.1
5.3.2 Material Properties Material Tungsten alloy Lead Stainless steel Carbon steel 5.4 Shielding Evaluation 5.4.1 Methods TABLE 5.3.1 SHIELDING PARAMETERS Density gm/cc 17 11.3 8.0 7.85 Mass Absorption Coefficient cm2/gm
.0555
.058
.054
.054 The methods of shielding calculations are described below. In addition, because the package has been in service for several years, the shielding evaluation can be informed and supported by shipment survey data. Table 5.4.1 presents dose rates for past shipments.
The first six examples use the Drawing 240122 item 4 configuration, while the last two use the item 5 configuration.
USA/9215/B(U) Consolidated Application for Renewal April 2018 Revision 6 Page 5-7
Source Distance Strength From Curies Package 8,050 Surface 1 meter 8,7004 Surface 1 meter 4,100 Surface 1 meter 9,soo8 Surface 1 meter 7,950 Surface 1 meter 7,300 Surface 1 meter 13,500 Surface 1 meter 6,5009 Surface 1 meter 5
Facing forward 6
E indicates estimate 7 At gamma background 8 Two source total 9
Pencil source FWD 2
0.3E 10 2
5
.2 4
1.5 9
1.5 4
0.9 25 max 5
TABLE 5.4.1 DOSE RATES FOR SHIPMENTS Maximum Dose Rate, mR/hr Back Left5 Right1 5
7 5
0.6 0.5 0.5 14 20 13 3
3 2
15 5
15 1
.5 1
5 18 30 2
5 5
11 25 25 1.5 8
5 9
20 18 1.5 3
3 85max 9
USA/9215/B(U) Consolidated Application for Renewal April 2018 Above 0.2
- 0. 1 0.6 0.4 GB7 GB3 0.8 0.6 0.1
.0. 3 0.7 0.4 Below 70E6 12E 60 8E 4.0
- 1. s 70E 9E 70 9
46 8
Revision 6 Page 5-8
From the Table 5.4.1, it can be seen that the 13,500 Ci source had a lower contact dose rate and a lower Tl than the 6,500 Ci source. This is due to geometry. The 13,500 Ci source was a teletherapy source, while the 6,500 Ci source was a longer cylindrical source. A higher activity source in that same elongated cylindrical geometry would likely have exceeded a Tl of 10, and would have been shipped exclusive use, in accordance with 10 CFR71.47 and 49 CFR 173.441.
5.4.1.1 Radial Gamma Attenuation. The specific shielding arrangement within the drawer or holder placed in the drum chamber may vary, However, for the purposes of this analysis, a 15,000 Ci cobalt-60 teletherapy source was used to represent the highest authorized package loading, from an energy standpoint10. A point source model was used. The attenuation from chamber wall to exterior of the inner container, [lo/l)s/Tc was taken as the product of the individual shielding components.;
[lo/lJs/Tc = eµn xn / Bn, where Bn is the buildup factor, µn the linear attenuation coefficient, Xn the thickness of the shield component under consideration, and n designates the particular shielding material component.
Table 5.4.1.1 lists the input parameters for the calculation, as well as the results.
The configuration is shown schematically in Figure 5.4.1.1. The constituent material attenuations are shown for each of the shielding component materials, as well as the total for both the Drawing 240122 Item 4 and Item 5 drawer configurations.
The ratio of the Item 5 to the Item 4 attenuation is 1.54. Looked at in another way, for the same surface dose, the Item 5 configuration would have to contain a source strength 54 percent greater. The Item 4 drum was not considered shielding limited at 9,500 curies, so that no absolute level of source strength can be determined by this means. However, this analysis is consistent with the activity limits of the authorized contents, with 9,500 Ci of cobalt-60 permitted for Item 4 and 15,000 Ci cobalt-60 permitted for Item 5.
The dose rate at the package surface and at one meter distant were also calculated in the radial direction. The attenuation due to shielding inside of the source containing drum chamber and the small attenuation due to the overpack were combined with the S/TC attenuation shown in Table 5.4.1.1 to provide the total material attenuation of the packaging. The additional constituents, as well as the overall result, are presented in Table 5.4.1.2. The overall shielding attenuation, 10 Due to the two photons associated with a cobalt-60 decay, and their higher energies, 15,000 Ci of cobalt-60 represents a significantly more difficult shielding challenge than 20,600 Ci of cesium-137.
USA/9215/B(U) Consolidated Application for Renewal April 2018 Revision 6 Page 5-9
[lo/I] is 5. 75 x 105. Combining this with the source dose relationship11 in the absence of shielding yields the following surface dose:
(5.2 x 106) (15,000)(2.5)/(5.75 x 105) (57.2}2= 104 mr./hr.
The dose at 1 meter is:
(5.2 x 106 x 15,000)(2.5)/(5.75 x 105) (157.2)2 = 14 mr./hr.
These values compare with 200 mR/hr and 10 mR/hr, respectively, for normal shipment. The margin provided for surface dose rate appears adequate for slight changes in shielding, thickness, geometry, or calculational uncertainty. To the extent the 10 mR/hr at 1 m dose rate is exceeded, the package would be shipped exclusive use, as described in 49 CFR 173.441. In actual practice, this model has proven to be conservative and generally overstates the actual dose rates. For example, from Table 5.4.1, the loading of 13,500 Ci actually only had a Tl of 5 and a contact dose rate of 25 mR/hr, compared with values predicted by this model of 13 and 94 mR/hr, respectively.
where 11 lo= Dose rate at distance d, cm from C curie source
= 5.2 x 106 CE/d2 mR/hr.
C = 15,000 curies E = Total gamma energy/disintegration = 2.5 MeV for cobalt-60 d (surface) - [(48.5/2)-1.75] 2.54 = 57.2 cm d (@ I meter)= 157.2 cm S. Glasstone, Principles of Nuclear Engineering, pg. 545 USA/9215/B(U) Consolidated Application for Renewal April 2018 Revision 6 Page 5-10
TABLE 5.4.1.1 CALCULATED RADIAL GAMMA ATTENUATION COMPARISON Location (1)
Material and Linear Drawing 240122 Thickness, in.
Absorption Item 4 / Item 5 Drum Coefficient cm-1 Configuration
- 1. Drum Liner s.s (3)
Item 5 0.095 0.432 Item 4 Same as above
- 2. Drum Shielding Lead Item 5 0.782 0.655 ltem4 0.625 0.655
- 3. Drum Casing S.S.
Item 5 o.187 0.432 ltem4 0.219 0.432
- 4. Shell Liner c.s. (3)
Item 5 0.187 0.424 ltem4 0.375 0.424
- 5. Shell Shielding Lead Item 5 7.69 0.655 ltem4 7.50 0.655 6.Shell c.s.
Item 5 0.375 0.424 Item 4 Same as above S/TC Attenuation, rr (Drawing 240122 Item 5)
S/TC Attenuation, rr (Drawing 240122 ltem 4)
Ratio, n {Item 5)/ rt (Item 4) locations shown in Figure 5.4.1.1 (1)
Numbers keyed to Buildup Factor (2) 1.09 1.47 1.40 1.17 1.20 1.17 1.34 4.75 4.65 1.34 Attenuation lo/I (2) 1.02 2.50 2.02 1-*****.*
1.046 1.056 1.044 7.63 X 104 5.68 X 104 2.37 X 105 1.538 X 103 1.54 (2)
Attenuation lo/I =eµnxn / Bn. Buildup factor based on point isotropic source.
Radiological Health Handbook, pgs. 145-146)
(3)
S.S.== stainless steel, C.S.== carbon steel.
USA/9215/B(U) Consolidated Application for Renewal April 2018 Revision 6 Page 5-11
- 6. SHELL 6
- 5. SHELL SHIELDING 5
- 4. SHELL LINER
- 4.
3
- 3. DRUM CASING
- 2.
- 2. DRUM SHIELDING I.
- 1. DRUM LINER s,
SOURCE CENTER Of DRUM CENTER FIG. 5.4.1.1 STRUCTURE & SHIELDING ARRANGEMENT KEY FOR TABLE 5.4.1.1 USA/9215/B(U} Consolidated Application for Renewal April 2018 Revision 6 Page 5-12
TABLE 5.4.1.2 CALCULATED DOSE RATE FOR DRAWING 240122 ITEM 5 PACKAGE Material And Linear Location Thickness in.
Absorption Coefficient, cm. -
1 Source capsule Tungsten alloy 0.437 0.944 Source chamber Stainless steel 0.314 0.432 Source chamber to Inner container From surface Table 5.4.1.1 Wooden Wood protective jacket 6.0 0.0384 Steel shell Carbon steel 0.107 0.424 Total material attenuation, source to package surface USA/9215/S{U) Consolidated Application for Renewal April 2018 Attenuation Buildup Factor
!/I 1.47 1.94 1.29 1.092 2.37 X 105 1.62 1.11 1.10 1.032 (lo/I) = 5. 75 x 105 Revision 6 Page 5-13
5.4.1.2 Axial Gamma Attenuation. Evaluation of the shielding in the direction parallel to the axis of the new inner container drum involves the source loading arrangement. The loading arrangement of a source in an international capsule is shown in Figure 4.3. This is representative and one of the more frequent types of loading arrangements. The 2.56 inch diameter drum chamber is fitted with a stainless steel sleeve having an outside diameter of 2.50 inches and an inside diameter of 2.060 inches. The capsule is placed within the sleeve and held in the axially central region of the drum with two tungsten alloy plugs, one on each side. The covers hold the entire assembly in place.
The arrangement in the Item 4 configuration is similar, except the drum chamber is 2.81 inches in diameter and a second sleeve of 0.095 wall thickness, surrounding the first is used to fill the space and center the source.
For both configurations the shielding arrangement in the axial direction is a plug of tungsten alloy 9.8 inches long and 2.03 inches in diameter (about twice the diameter of the source face) surrounded by an annulus of steel with either two or three narrow air gaps. This assembly, in turn, is surrounded by a matrix of lead. The arrangement is shown for the Item 5 configuration in Figure 5.4.2.1.
Based on a point source, a simple calculation shows that for a shield thickness of 9.8 inches (the length of the plug and approximate distance from the source to the face of the shell assembly), the attenuation in tungsten alloy is on the order of 10 10, that in lead of order 107, and in steel of order 104* With the highest leakage path being that through the annulus of steel, a comparative measure of attenuation between the Item 5 and the Item 4 arrangements can be made by treating the steel annulus as a streaming path. The annulus is thinner in the Item 5 arrangement. To determine the relative streaming, the following expression(1l for the ratio of entering to leaving gamma flux was used and taken as proportional to the corresponding dose rates:
( 1) Source: T. Rockwell, Reactor Shielding Manual, pg. 293 USA/9215/B(U) Consolidated Application for Renewal April 2018 Revision 6 Page 5-14
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TUNGSTEN ALLOY STAINLESS STEEL C
9.s,
.9 t--
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(.O40m. min. S.S.)
~ER
'-~GASKET YI
- LEAD
~
C. STEEL fe to1/2 SILICONE GASKET
10.81-----'------
FIG. 5.4.2.l DETAIL OF INTERNATIONAL CAPSULE POSITIONING WITHIN INNER CONTAINER DRUM.
( REF. FIG. 4.3 )
¢ / ¢1 =1/2re L2 [(cos-1 r/R) (2R2 - r2) - r(R2 - r2) 1/2J The definition of the symbols and the corresponding values for both the Item 5 and Item 4 configurations used in the comparisons are as follows:
¢, gamma flux (taken proportional to dose rate}
R, drum chamber radius, in.
R, shield plug radius, in.
L, comparative shield thickness, in.
Forthe Item 4 units:
Item 4 S/TC's 1.405 1.02 9.81
¢ / ¢0 = 1/2n(9.81)2[(cos-1 1.02/1.405) {2(1.405)2 - (1.02)2) 102((1.405)2-(1.02}2)112]
= 2.01 X 10-3 For the Item 5 units:
¢ / ¢ 0 = 1/2n(9.81)2[(cos-1 1.02/1.28) (2(1. 78)2- (1.02)2) 102((1.28}2 - (1.02)2)112]
= 1.095 X 10*3 Item 5 S/TCs 1.280 1.02 9.81 The increase in attenuation is proportional to 2.01/1.095 or 1.84, which is close to a factor of two.
Another loading arrangement that occurs frequently is one in which the entire teletherapy machine drawer, with source loaded, is carried in the drum chamber. In the case of the AECL/Theratronics/Best machine, for example, the shielded drawer, with center positioned source, is the full length of the drum chamber and 2.475 inches in diameter. The input value for the calculation are:
USA/9215/B(U) Consolidated Application for Renewal April 2018 Revision 6 Page 5-16
R R
L Substituting the new values:
For the Item 4 units For the Item 5 units Item 4 S/TC's 1.405 1.234 10.8
I <l>o = 5.187 X 10*4
¢, / $0 = 7.01 X 10-5 Item 5 S/TC's 1.280 1.234 10.8 The increase in attenuation for the new units is 5.187 /.701 = 7.3 or a factor of about seven.
Other specific cases will vary, but the difference between the two configurations iis significant, again justifying the higher activity limit for the Item 5 configuration.
5.4.2 Input and Output Data The input and output data, along with analytical examples and demonstration of convergence has been included in Section 5.4.1.
5.4.3 Flux to Dose Rate Conversion In addition to the formula given above in 5.4.1:
1 Ci of cobalt-60 gives a dose rate of 1.3 R/hr at 1 meter; and 1 Ci of Cs-137 gives a dose rate of 0.35 R/hr at 1 meter.
5.4.4 External Radiation Levels Table 5.4.1 provides several different loading scenarios, and the associated dose rates. The analysis in 5.4.1 demonstrates that the shielding models used produce very conservative results when compared to actual measurements. This section provides another specific example.
USA/9215/B(U) Consolidated Application for Renewal April 2018 Revision 6 Page 5-17
The results of a radiation survey of a package incorporating the Item 5 configuration are provided in this section. The survey was made on December 4, 1986, on a package that had been prepared for shipment and sealed a few days before. The source strength was 6,650 curies (12/1/86). The source was fitted into an international capsule and held in the central region of the drum chamber between tungsten alloy end plugs. The remaining drum chambers were loaded with full length, lead filled plugs. Measurements were made with a calibrated G-M detector.
The package surface measurements are shown in Figure 5.4.3.1. All of the radiation entries are in mR/hr The maximum reading was 15 mR/hr at the center of the package bottom.
The highest side readings were 14 mR/hr and 6 mR/hr, located 180° from each other at a belt line height of 24 inches. The remaining surface readings were between 0.6 and 5 mR/hr at locations as shown on Figure 5.4.3.1.
All readings taken at one meter distant from the package surface were 1 mR/hr or less. No measurement was taken at one meter beneath the bottom of the package.
Based on these measurements, the design basis 15,000 curie source would result in a maximum surface reading of (15,000/6,650) X 15 = 34 mR/hr, as compared with the design basis value of 100 mR/hr and the 10 CFR 71.47 limit of 200 mR/hr These results also generally support the calculations provided in Section 5.4.1, and are consistent with the data in Table 5.4.1, particularly in comparison with the 13,500 Ci source which had a maximum surface reading of 25 mR/hr.
Table 5.4.4.1 applies the shielding evaluation methods of this section to the loading depicted in Figure 5.3.1, focusing on the area of the expected highest dose rate for this configuration, and compares the results with the survey data shown in Figure 5.4.3.1.
USA/9215/B{U} Consolidated Application for Renewal April 2018 Revision 6 Page 5-18
TABLE 5.4.4.1 CALCULATED DOSE RATE FOR THE LOADING CONFIGURATION DEPICTED IN FIGURE 5.3.1 Location (1)
Material and UnearAbsorption Buildup Attenuation Figure 5.4.2.2 Thickness, in.
Coefficient cm-1 Factor (2) lo/I (2)
- 1. Co-60 source -
cobalt self attenuation 0.80 0.482 1.67 1.6 2.Shield Plug Tungsten alloy 3.12 0.944 3.5 510
- 3. Sleeve and liner Stnls. Steel 1.65 0.432 2.25 2.7
- 4. Drum Lead 2.55 28 0.655 2.5 5.Drum casing+
Steel shell casing+ flange 3.45 0.432 3.6 12 S/TC Attenuation, rr (Axial shielding for Figure 5.4.2.2}
7.4 X 105 The distance from the center of the source to the surface of the S/TC steel flange is approximately 11.5". The dose rate from a 6,650 Ci source at the S/TC surface would be:
[(6,650 Ci}(l.3 R/hr)(l/11.5 x 0.0254)2] / 7.4 x 105== 141 mR/hr The surface of the overpack in the highest dose rate area would be approximately 25" from the center of the source. Using the inverse square law, and not taking into account any shielding benefit from either the WPJ or the Steel Shell, gives a maximum surface dose of:
(141 mR/hr) (11.5/25}2 = 30 mR/hr.
USA/9215/B(U) Consolidated Application for Renewal April 2018 Revision 6 Page 5-19
Again, this is a conservative estimate for this loading, which resulted in a measured maximum surface dose rate of 14 mR/hr.
Any change in shielding effect resulting from the Hypothetical Accident Conditions is due to shifting of the inner container within the overpack. There is no opportunity for any measurable shift of source or shielding in the S/TC under the most severe free drop and fire conditions. Except for some small clearances, the S/TC is completely filled with metal.
The maximum shift of the S/TC within the overpack can be obtained from the analysis of the several hypothetical accident drop conditions. The shift of the source relative to the outer surface of the package is due to the crushing, bending, or other distortion of the overpack wooden protective jacket (WPJ) and steel shell (SS). The results obtained from the accident analysis are summarized in the following table:
USA/9215/B(U) Consolidated Application for Renewal April 2018 Revision 6 Page 5-20
Maximum Displacement of Source Relative to Normal location in Component of Packages1 inches Overpack Top Bottom Side Edge Affected Drop Drop Drop Drop(12)
Crush support beams (SS) 4 Shred shock rings (WPJ) 2 Inner container movement 1
2 Inner container penetration of WPJ 4
1 After fire drop, char allowance
.?.
l
.f.
Maximum Displacement, in.
6 7
7 12 Not critical for shielding The amount of shielding material will remain the same. The shielding change will result only from geometric factors. Postulating a point isotopic source, the increased transmission due to the seven inch maximum displacement is:
At the surface:
At one meter:
(~2==1.97
( 24.4-7)2
(§A._)2 = 1.26
( 63.8-7)2 Assuming the surface radiation level under pre-accident conditions was at the l00mR/hr design basis condition, the hypothetical accident would result in a surface radiation level of less than 200 mR/hr Similarly, postulating the permissible 10 mR/hr pre-accident Tl, the post-accident one meter dose rate increase would be less than 3 mR/hr In any case, both levels are below the 10 CFR71.51(a)(2) limit of 1 rem/hr at one meter from the external surface of the package under hypothetical accident conditions.
USA/9215/B(U) Consolidated Application for Renewal April 2018 Revision 6 Page 5-21
0.6 TO 1.0 3 TO 4.5 241n.
l FAR SIDE NOTE : FAR SIDE READING IN (
)
/J~L READINGS IN m,;h,.
2.5 TO 3 2 TO 3 \\'
s-------+ ---,------2.3 +
0.6 15 0.6 TO 2 (O.6T0 2)
O.6T03 (3}
FIG. 5.4.3.I UPPER PERIPHERY OB ALL AROUND 1.0 2.3 LOWER PERIPHERY
<3 ALL AROUND EXCEPT AS NOTED MEASURED PACKAGE SURFACE RADIATION LEVELS
. FOR A CONTAINED SOURCE OF 6650 CURIES Revision 1 5-15 USA/9215/B(U) Consolidated Application for Renewal April 2018 Revision 6 Page 5-22